In this work the authors solved the steady state neutron diffusion equation for a multi-layer slab assuming the multi-group energy model. The method to solve the equation system is based on a expansion in Taylor Series, which was proven to be useful in [1] [2] [3]. The results obtained can be used as initial condition for neutron space kinetics problems. The neutron scalar flux was expanded in a power series, and the coefficients were found by using the ordinary differential equation and the boundary and interface conditions. The effective multiplication factor k was evaluated using the power method [4]. We divided the domain into several slabs to guarantee the convergence with a low truncation order. We present the formalism together with some numerical simulations. (author)

In this work the authors solved the steady state neutron diffusion equation for a multi-layer slab assuming the multi-group energy model. The method to solve the equation system is based on an expansion in Taylor Series resulting in an analytical expression. The results obtained can be used as initial condition for neutron space kinetics problems. The neutron scalar flux was expanded in a power series, and the coefficients were found by using the ordinary differential equation and the boundary and interface conditions. The effective multiplication factor k was evaluated using the power method. We divided the domain into several slabs to guarantee the convergence with a low truncation order. We present the formalism together with some numerical simulations.

Neutronfluxes for energies between 15 MeV and thermal at depths of 0 to 300 g/sq cm in the moon are calculated by the discrete ordinate mathod with the ANISN code. With the energy spectrum of Lingenfelter et al. (1972). A total neutron-production rate for the moon of 26 plus or minus neutrons/sq cm sec is determined from the Ar-37 activity measurements in the Apollo 16 drill string, which are found to have a depth dependence in accordance with a neutron source function that decreases exponentially with an attenuation length of 155 g/sq cm.

Neutronflux monitor (NFM) as an important diagnostic sub-system in ITER (international thermonuclear experimental reactor) provides a global neutron source intensity, fusion power and neutronflux in real time. Three types of neutronflux monitor assemblies with different sensitivities and shielding materials have been designed. Through MCNP (Mante-Carlo neutral particle transport code) calculations, this extended system of NFM can detect the neutronflux in a range of 104 n/(cm2·s) to 1014 n/(cm2·s). It is capable of providing accurate neutron yield measurements for all operational modes encountered in the ITER experiments including the in-situ calibration. Combining both the counting mode and Campbelling (MSV; Mean Square Voltage) mode in the signal processing units, the requirement of the dynamic range (107) for these NFMs and time resolution (1 ms) can be met. Based on a uncertainty analysis, the estimated absolute measurement accuracies of the total fusion neutron yield can reach the required 10% level in both the early stage of the DD-phase and the full power DT operation mode. In the advanced DD-phase, the absolute measurement accuracy would be better than 20%.

Using Indium activation foils, the toroidal and poloidal neutron emission patterns were determined for PLT plasmas which include ICRF and neutral beam heating. The activities produced the /sup 115/In (n,n') /sup 115m/In reaction were determined by counting the 336 keV ..gamma.. line of the /sup 115m/In decay. This activation cross section falls just below 2.5 MeV so that the influence of scattered neutrons of degraded energies is reduced. From the magnitude of the activity, the absolute calibration of the PLT fusion neutron emission is obtained with less than or equal to 40% accuracy.

A 14-MeV neutron source based on a Gas-Dynamic Trap will provide a high flux of 14 MeV neutrons for fusion materials and sub-component testing. In addition to its main goal, the source has potential applications in condensed matter physics and biophysics. In this report, the author considers adding one more capability to the GDT-based neutron source, the modulation of the neutronflux with a desired frequency. The modulation may be an enabling tool for the assessment of the role of non-steady-state effects in fusion devices as well as for high-precision, low-signal basic science experiments favoring the use of the synchronous detection technique. A conclusion is drawn that modulation frequency of up to 1 kHz and modulation amplitude of a few percent is achievable. Limitations on the amplitude of modulations at higher frequencies are discussed.

A simple and easy method for measuring the neutronflux is presented. This paper deals with the experimental verification of neutron dose rate–flux relationship for a non-dissipative medium. Though the neutronflux cannot be obtained from the dose rate in a dissipative medium, experimental result shows that for non-dissipative medium one can obtain the neutronflux from dose rate. We have used a 241 AmBe neutron source for neutron irradiation, and the neutron dose rate and count rate were measured using a NM2B neutron monitor and R-12 superheated droplet detector (SDD), respectively. Here, the neutronflux inferred from the neutron count rate obtained with R-12 SDD shows an excellent agreement with the flux inferred from the neutron dose rate in a non-dissipative medium.

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutronflux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutronflux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutronflux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutronflux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutronflux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

Neutronflux density measurement by thermal analysis is a new method different from the previous.This method is first put the sample to the neutron field.Second,measure the annealingheat of the sample.Find out the suitable mixture of crystal boron and apatite to measure the neutronflux density.Then put the sample to the neutron field in

On Dec. 3, 1996, the Swiss spallation neutron source SINQ was taken into operation and produced its first neutrons successfully. The neutron spectrum within one of the supermirror guides was estimated by a chopper Time-of-Flight method. The result shows a 30% higher neutron intensity at the flux maximum than expected from previous Monte-Carlo simulations. (author) 1 fig., 4 refs.

The measurement and calculation of the neutronflux mapping of the OPAL research reactor are presented. Following an investigation of fuel coolant channels using sub-miniature fission chambers to measure thermal neutronflux profiles, neutronic calculations were performed. Comparison between calculation and measurement shows very good agreement.

The problem of neutronflux measurements using activation detectors is studied in the particular case of protection research. It is shown how it possible, it is possible, using a known thermal flux, to organise a coherent calibration system for all the detectors. The rapid neutron detectors are calibrated with respect to a reference detector (phosphorus) in a natural uranium converter; the intermediate neutron detectors with respect to gold in the axial channel of ZOE. This method makes it possible to minimise the errors due to the activation cross-sections. A brief description is given of the counting room of the Pile Safety Study Service, as well of the practical utilisation characteristics of the counters employed. (authors) [French] Le probleme de la mesure des flux de neutrons au moyen de detecteurs par activation est etudie dans le cas particulier des etudes de protections. On montre comment, a partir d'un flux thermique connu, on peut organiser un systeme coherent d'etalonnage de tous les detecteurs. Les detecteurs de neutrons rapides sont etalonnes par rapport a un detecteur de reference (phosphore) dans un convertisseur en uranium naturel; les detecteurs de neutrons intermediaires, par rapport a l'or dans le canal axial de ZOE, Cette methode permet de minimiser les erreurs dues aux sections efficaces d'activation. On decrit sommairement la salle de comptage du Service d'Etudes de Protections de Piles et on indique les caracteristiques d'emploi pratique des detecteurs utilises. (auteurs)

The results of research of a thermalizer (heater) width influence on neutron thermalization efficiency during FCM neutronflux measuring in the 'Ukryttia' are described. The calculations of neutronflux densities were performed by the Monte-Carlo method with the help of computer code MCNP-4C for FCM different models.Three possible installations of detectors were considered: on FCM surface,inside the FCM, and inside the concrete under the FCM layer. It was shown,that in order to increase the sensitivity of neutron detectors in intermediate and fast neutrons field,and consequently, to decrease the dependence of the readings of spectral distribution of neutronflux,it is necessary to position the detector inside the so-called thermalizer or heater. The most reasonable application of thick 'heaters' is the situation, when the detector is placed on FCM surface.

The prototype neutronflux monitor consists of a high purity 235U fission chamberdetector and a "blank" detector, which is a fissile material free detector with the same dimensionas the fission chamber detector to identify noise issues such as noise coming from gamma rays. Themain parameters of the fission chamber assembly that have been measured in the laboratory areconfirmed to approach the technological level of the International Thermonuclear ExperimentalReactor (ITER) in the near future. This prototype neutronflux monitor will be further developedto become a neutronflux monitor suitable for the operation phase of D-D fusion on the ITER.

Full Text Available In a seismically active zone in the near Almaty area (Kazakhstan since 1996 observations of variations of a natural neutronflux have been conducted. Sometimes the neutronflux rises sharply within the one-hour interval in comparison with the background. It occurs on the eve of activation of seismic processes. Increase of the neutronflux level had taken place from 1 h to 10 days prior to earthquakes. It is also indicated a tendency of growth of the anomaly level in accordance with the growth of energetic class of the subsequent earthquake. A character of connection between the neutronflux and earthquakes is still not clear. It is proposed that the neutronflux anomalies caused by variations of cosmic radiation intensity under action of fluxes of solar material, which is burst into interplanetary space (solar wind during solar flares. Energy of the solar wind transferred to Earth puts into action a trigger mechanism of the process of initiation of earthquakes at those places where conditions have already been prepared for them. The neutronflux anomalies can be used as substantial additional information for classical geophysical methods of short-term earthquake prediction.

Investigation of subthermonuclear plasma confinement and heating in magnetic fusion devices such as GOL-3 and GDT at the Budker Institute (Novosibirsk, Russia) requires sophisticated equipment for neutron-, gamma- diagnostics and upgrading data acquisition systems with online data processing. Measurement of fast neutronflux with stilbene scintillation detectors raised the problem of discrimination of the neutrons (n) from background cosmic particles (muons) and neutron-induced gamma rays (γ). This paper describes a fast neutronflux analyzer with real-time digital pulse-shape discrimination (DPSD) algorithm FPGA-implemented for the GOL-3 and GDT devices. This analyzer was tested and calibrated with the help of 137Cs and 252Cf radiation sources. The Figures of Merit (FOM) calculated for different energy cuts are presented.

It is known that neutron background is a major problem for low-background experiments carrying out underground, such as dark matter search, double-beta decay searches and other experiments known as Underground Physics. We present here some results obtained with the en-detector of 0.75 m2, which is running for more than 4 years underground at a depth of 25 m water equivalent in Skobeltsyn Institute of Nuclear Physics, Moscow State University. Some spontaneous increases in thermal neutronflux up to a factor of 3 were observed in delayed anti-correlation with barometric pressure. The phenomenon can be explained by the radon barometric pumping effect resulting in similar effect in neutronflux being produced in (α, n)-reactions by alpha-decays of radon and its daughters in surrounding rock. This is the first demonstration of the barometric pumping effect observed in thermal neutronflux underground.

Americium–beryllium (Am-Be; n, γ) is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci), yielding a small thermal neutronflux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources) experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutronflux compared to the traditional method with water moderator.

With the development of innovative nuclear systems and new generation neutron sources, the nuclear instrumentation should be adapted. Since several years, we developed microscopic fission chambers to study the transmutation of minor actinides in high thermal-neutronfluxes. The recent developments done to fulfill the drastic conditions of irradiations are described in this paper together with the feedback from the measurements. Two installations were used: the HFR of the ILL for its highest thermal neutronflux of the world and the MEGAPIE target which was the first 1 MW liquid Pb-Bi spallation target in the world. (authors)

Neutron generators based on inertial electrostatic confinement fusion were considered for the design of a neutron irradiation facility for explanted organ Boron Neutron Capture Therapy (BNCT) that could be installed in a health care center as well as in research areas. The chosen facility configuration is "irradiation chamber", a ~20×20×40 cm(3) cavity near or in the center of the facility geometry where samples to be irradiated can be placed. Neutronflux calculations were performed to study different manners for improving scattering processes and, consequently, optimize neutronflux in the irradiation position. Flux distributions were assessed through numerical simulations of several models implemented in MCNP5 particle transport code. Simulation results provided a wide spectrum of combinations of net fluxes and energy spectrum distributions. Among them one can find a group that can provide thermal neutronfluxes per unit of production rate in a range from 4.1·10(-4) cm(-2) to 1.6·10(-3) cm(-2) with epithermal-to-thermal ratios between 0.3% and 13% and fast-to-thermal ratios between 0.01% to 8%. Neutron generators could be built to provide more than 10(10) n s(-1) and, consequently, with an arrangement of several generators appropriate enough neutronfluxes could be obtained that would be useful for several BNCT-related irradiations and, eventually, for clinical practice.

Optimization of neutronfluxes in experimental channels is of great concern in research reactor utilization. The general approach used at the NUR research reactor for neutronflux optimization in irradiation channels is presented. The approach is essentially based upon a judicious optimization of the core configuration combined with the improvement of reflector characteristics. The method allowed to increase the thermal neutronflux for radioisotope production purposes by more than 800%. Increases of up to 60% are also observed in levels of useful fluxes available for neutron diffraction experiments (small angle neutron scattering (SANS), neutron reflectometry, etc.). Such improvements in the neutronic characteristics of the NUR reactor opened new perspectives in terms of its utilization. More particularly, it is now possible to produce at industrial scales major radio-isotopes for medicine and industry and to perform, for the first time, material testing experiments. The cost of the irradiations in the optimized configuration is generally small when compared to those performed in the old configuration and an average reduction factor of about of 10 is expected in the case of production of Molybdenum-99 (isotope required for the manufacturing of Technetium-99 medical kits). In addition to these important results, safety analysis studies showed that the more symmetrical nature of the core geometry leads to a more adequately balanced reactivity control system and contributes quite efficiently to the operational safety of the NUR reactor. Results of comparisons between calculations and measurements for a series of parameters of importance in reactor operation and safety showed good agreement.

The Los Alamos Isotope Production Facility (IPF) creates medical isotopes using a proton beam impinged on a target stack. Spallation neutrons are created in the interaction of the beam with target. The use of these spallation neutrons to produce additional radionuclides has been proposed. However, the energy distribution and magnitude of the flux is not well understood. A modified SAND-II spectral adjustment routine has been used with radioactivation foils to determine the differential neutron fluence for these spallation neutrons during a standard IPF production run.

The Los Alamos Isotope Production Facility (IPF) creates medical isotopes using a proton beam impinged on a target stack. Spallation neutrons are created in the interaction of the beam with target. The use of these spallation neutrons to produce additional radionuclides has been proposed. However, the energy distribution and magnitude of the flux is not well understood. A modified SAND-II spectral adjustment routine has been used with radioactivation foils to determine the differential neutron fluence for these spallation neutrons during a standard IPF production run.

Realizations of the TOF technique for neutron reflectometers at steady flux reactors are compared. Beam choppers for neutron reflectometers divide into choppers of type 1 (Δλ = const) and 2 (Δλ / λ = const) . It follows from Monte-Carlo simulations that choppers of type 1 do not yield to more intricate choppers of type 2, widely used at neutron reflectometers. Because of a very fast drop of neutron reflectivities with the momentum transfer q, non-optimality of measurements with a chopper of type 1 is fully compensated by better statistics at large q, and is not so much essential at small q. To vary the TOF resolution with choppers of type 1, a phasing of two discs and a turning of the system of two discs are suggested. The fluxes of neutrons with wavelengths beyond the working range and the efficiencies of their elimination by means of a bandwidth limiting prechopper are evaluated.

The latest neutron spectrometer measurements with the Lunar Exploration Neutron Detector (LEND) onboard the Lunar Reconnaissance Orbiter (LRO) are presented. It covers more than 1 year of mapping phase starting on 15 September 2009. In our analyses we have created global maps showing regional variations in the flux of thermal (energy range 0.5 MeV), and compared these fluxes to variances in soil elemental composition, and with previous results obtained by the Lunar Prospector Neutron Spectrometer (LPNS). We also processed data from LEND collimated detectors and derived a value for the collimated signal of epithermal neutrons based on the comparative analysis with the LEND omnidirectional detectors. Finally, we have compared our final (after the data reduction) global epithermal neutron map with LPNS data.

The conformity of the mock-up results and those for reactor pressure vessel (RPV) of nuclear power plants (NPP) has been evaluated in order to qualify if the mock-ups data could be used for benchmark's purpose only, or/and for simulating of the NPP irradiation conditions. Neutron transport through the vessel has been calculated by the three-dimensional discrete ordinate code TORT with problem oriented multigroup energy neutron cross-section library BGL. Neutronflux/fluence and spectrum shape represented by normalized group neutronfluxes in the multigroup energy structure, for neutrons with energy above 0.5 MeV, have been used for conformity analysis. It has been demonstrated that the relative difference of the attenuation factor as well as the group neutronfluxes did not exceed 10% at all considered positions for VVER-440. For VVER-1000, it has been obtained the same consistency, except for the location behind the RPV. The neutronflux attenuation behind the RPV is 18% higher than the mock-up attenuation. It has been shown that this difference arises from the dissimilarity of the biological shielding. The obtained results have demonstrated that the VVERs' mock-ups are appropriate for simulating the NPP irradiation conditions. The mock-up results for VVER-1000 have to be applied more carefully i.e. taking into account the existing peculiarity of the biological shielding and RPV attenuation azimuthal dependence.

The objective of this work is to implement various fast neutronflux reduction programs on the belt-line region of the reactor pressure vessel to reduce the increasing rate of reference temperature for pressurized thermal shock (RT PTS) for Korea Nuclear Unit 1. A pressurized thermal shock (PTS) event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in the belt-line region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Generally, the RT PTS value is continuously increasing according to the fast neutron irradiation during the reactor operation, and it can reach the screening criterion prior to the expiration of the operating license. To reduce the increasing rate of RT PTS, various neutronflux reduction programs can be implemented, which are focused on license renewal. In this paper, neutronflux reduction programs, such as low leakage loading pattern strategy, loading of neutron absorber rods, and dummy fuel assembly loading are considered for Korea Nuclear Unit 1, of which the RT PTS value of the leading material (circumferential weld) is going to reach the screening criterion in the near future. To evaluate the effects of the neutronflux reduction programs, plant and cycle specific forward neutron transport calculations for the various neutronflux reduction programs were carried out. For the analysis, all transport calculations were carried out by using the DORT 3.1 discrete ordinate code and BUGLE-96 cross-section library. (authors)

The upper limit of the density of the thermal neutronflux from pulsed sources based on the fission reaction is established. Three types of sources for research on ejected beams are considered: a multiplying target of the proton accelerator (a booster), a booster with the reactivity modulation (a superbooster), and a pulsing reactor. Comparison with other high-flux sources is carried out. The investigation has been performed at the Frank Laboratory of Neutron Physics of JINR.

It is known that neutron background is a big problem for low-background experiments in underground Laboratories. Our global net of en-detectors sensitive to thermal neutrons includes the detectors running both on the surface and at different depths underground. We present here results obtained with the en-detector of 0.75 m^2 which is running more than 3 years in underground room at a depth of 25 m of water equivalent in Skobeltsyn Institute of Nuclear Physics, Moscow. Spontaneous increases in thermal neutronflux up to a factor of 3 were observed in delayed anti-correlation with barometric pressure. The phenomenon can be explained by a radon barometric pumping effect resulting in similar effect in neutronflux produced in (alpha,n)-reactions by alpha-decays of radon and its daughters in surrounding rock

PhoNeS (photo neutron source) is a project aimed at the production and moderation of neutrons by exploiting high energy linear accelerators, currently used in radiotherapy. A feasibility study has been carried out with the scope in mind to use the high energy photon beams from these accelerators for the production of neutrons suitable for boron neutron capture therapy (BNCT). Within these investigations, it was necessary to carry out preliminary measurements of the thermal neutron component of neutron spectra, produced by the photo-conversion of X-ray radiotherapy beams supplied by three LinAcs: 15 MV, 18 MV and 23 MV. To this end, a simple passive thermal neutron detector has been used which consists of a CR-39 track detector facing a new type of boron-loaded radiator. Once calibrated, this passive detector has been used for the measurement of both the thermal neutron component and the cadmium ratio of different neutron spectra. In addition, bubble detectors with a response highly sensitive to thermal neutrons have also been used. Both thermal neutron detectors are simple to use, very compact and totally insensitive to low-ionizing radiation such as electrons and X-rays. The resultant thermal neutronflux was above 10{sup 6} n/cm{sup 2}s and the cadmium ratio was no greater than 15 for the first attempt of photo-conversion of X-ray radiotherapy beams.

A zero power multiplying assembly in subcritical regime serves as a benchmark for validating subcritical reactor physics. The utilization of a subcritical assembly for the determination of nuclear parameters in a multiplying medium requires a well-defined neutronflux to carry out the experiments. For this it is necessary to know the neutronflux profile inside a subcritical system. A compact subcritical assembly BRAHMMA has been developed in India. The experimental channels in this assembly are typically less than 8 mm diameter. This requires use of miniature detectors that can be mounted in these experimental channels. In this article we present the thermal neutronflux profile measurement in a compact subcritical system using indigenously developed miniature gas filled neutron detectors. These detectors were specially designed and fabricated considering the restrictive dimensional requirements of the subcritical core. Detectors of non-standard size with various sensitivities, from 0.4 to 0.001 cps/nv were used for neutronflux of interest ranging from 103 to 107 n-cm-2 s-1. A comparison of measured neutronflux using these detectors and simulated Monte Carlo calculations are also presented in this article.

Full Text Available The procedures for the numerical and experimental determination of the neutronflux in the zones with the strong neutron absorption and leakage are described in this paper. Numerical procedure is based on the application of the SCALE-4.4a code system where the Dancoff factors are determined by the VEGA2DAN code. Two main parts of the experimental methodology are measurement of the activity of irradiated foils and determination of the averaged neutron absorption cross-section in the foils by the SCALE-4.4a calculation procedure. The proposed procedures have been applied for the determination of the neutronflux in the internal neutron converter used with the RB reactor core configuration number 114.

The absolute value of the neutronflux is of paramount importance in reactor physics and other application on the nuclear field. Due to several corrections which should be done, such as radioactive decay of the produced nuclides, normalization factors between different irradiations, neutron spectrum perturbation, cross section behaviour and growing of the reactor power, among other factors, make the calculation of the neutronflux very cumbersome. the software FLUXO was developed to overcome these inconveniences. It is programmed in FORTRAN language, and was written to calculate the absolute flux of thermal, epithermal and fast neutrons, through the foil activation technique. The magnitude of this activation can be measured by a 4{pi} {beta}-{gamma} coincidence measurement or by gamma spectroscopy alone. The software calculates as well, the absolute activity of radioactive sources, and reactor-irradiated samples. (author)

G. Srinivasan et al. (1990) proposed a simple and elegant explanation for the reduction of the neutron star magnetic dipole moment during binary evolution leading to low mass X-ray binaries and eventually to millisecond pulsars: Quantized vortex lines in the neutron star core superfluid will pin against the quantized flux lines of the proton superconductor. As the neutron star spins down in the wind accretion phase of binary evolution, outward motion of vortex lines will reduce the dipole magnetic moment in proportion to the rotation rate. The presence of a toroidal array of flux lines makes this mechanism inevitable and independent of the angle between the rotation and magnetic axes. The incompressibility of the flux-line array (Abrikosov lattice) determines the epoch when the mechanism will be effective throughout the neutron star. Flux vortex pinning will not be effective during the initial young radio pulsar phase. It will, however, be effective and reduce the dipole moment in proportion with the rotation rate during the epoch of spindown by wind accretion as proposed by Srinivasan et al. The mechanism operates also in the presence of vortex creep.

The nuclear design of fusion devices such as ITER (International Thermonuclear Experimental Reactor), which is an experimental fusion reactor based on the "tokamak" concept, rely on the results of neutron physical calculations. These depend on the knowledge of the neutron and photon flux spectra which is particularly important because it permits to anticipate the possible answers of the whole structure to phenomena such as nuclear heating, tritium breeding, atomic displaceme...

The procedures for the numerical and experimental determination of the neutronflux in the zones with the strong neutron absorption and leakage are described in this paper. Numerical procedure is based on the application of the SCALE-4.4a code system where the Dancoff factors are determined by the VEGA2DAN code. Two main parts of the experimental methodology are measurement of the activity of irradiated foils and determination of the averaged neutron absorption cross-section in the foils by t...

The High FluxNeutron Generator (HFNG) is a DD plasma-based source, with a self-loading target intended for fundamental science and engineering applications, including 40 Ar/39 Ar geochronology, neutron cross section measurements, and radiation hardness testing of electronics. Our first estimate of the neutron yield, based on the population of the 4.486 hour 115 In isomer gave a neutron yield of the order 108 n/sec; optimization is ongoing to achieve the design target of 1011 n/sec. Preliminary x-ray spectra showed prominent energy peaks which are likely due to atomic line-emission from back-streaming electrons accelerated up to 100 keV impinging on various components of the HFNG chamber. Our x-ray and neutron diagnostics will aid us as we continue to evolve the design to suppress back-streaming electrons, necessary to achieve higher plasma beam currents, and thus higher neutronflux. This talk will focus on the characterization of the neutron yield and x-ray spectra during our tests. A collimation system is being installed near one of the chamber ports for improved observation of the x-ray spectra. This work is supported by NSF Grant No. EAR-0960138, U.S. DOE LBNL Contract No. DE-AC02-05CH11231, U.S. DOE LLNL Contract No. DE-AC52-07NA27344, and the UC Office of the President Award 12-LR-238745.

A new fusion experiment, the WENDELSTEIN 7-X Stellarator (W7-X), will be undertaken in Greifswald in Germany. Measurements of the neutronflux will provide information on fusion reaction rates and possibly also on ion temperatures as function of time. For this purpose moderating neutron counters will be designed, tested, calibrated and eventually used at W7-X. Extensive Monte-Carlo simulations have been performed in order to select the most suitable detector and moderator combination with a flat response function and highest achievable efficiency. Different detector configurations with different moderating materials have been tried out, showing that a 32 cm thick graphite moderating BF{sub 3} -counter gives the desired flat response and sufficient efficiency. Neutron spectra calculations have been made for different torus models and the influence of floor, walls and ceiling (i.e. reactor hall) have been investigated. Presented results suggest that a more detailed torus model significantly reduces the number of neutron counts at the detector. Calculations including the reactor hall indicate a tendency of shifting the neutron spectra towards the thermal region. The main part of the scattered neutrons are back-scattered from the floor. Finally, calculations on the graphite moderating BF{sub 3} -counter in the detailed torus environment were performed in order to assess the absolute response function under the influence of the reactor hall. The results show that the detector count rate will increase by only 5-7 % when the reactor hall is taken into account. With a stellarator generating 10{sup 12} to 10{sup 16} neutrons per second the detector count rate will be 2x10{sup 5} to 2x10{sup 9} neutrons per second.

The attenuation of the thermal -neutronflux is increased by adding borax. The neutron dose was reduced by a factor of approximately four by a...the thermal -neutronflux is increased by adding borax. The neutron dose was reduced by a factor of approximately four by a concrete box three feet on a...the ,, eutrons and their spatial distribution is of basic importance to the assessment of the effects of the neutrons from a device. Measurements of this

The Pu-239 fission chambers are widely used to measure fission spectrum neutronflux due to a flat response to fast neutrons. However, in the mean time the resonance and thermal neutrons can cause a significant influence on the measurement if they are moderated, which could be eliminated by using B and Cd covers. At a column enriched uranium fast neutron critical assembly, the fission reaction rates of Pu-239 are measured as 1.791*10-16,2.350*10-16 and 1.385*10-15 per second for 15mm thick B cover, 0.5mm thick Cd cover, and no cover respectively. While the fission reaction rate of Pu-239 is rapidly increased to 2.569*10-14 for a 20mm thick polythene covering fission chamber. The average Pu-239 fission cross-section of thermal and resonance neutrons is calculated to be 500b and 24.95b with the assumption of 1/v and 1/E spectra respectively, then thermal, resonance and fast neutronflux are achieved to be 2.30*106,2.24*106 and 1.04*108cm-2s-1.

The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a) to provide a thermal neutronflux in the neutron transmutation silicon doping, (b) to provide a thermal flux in the neutron activation analysis position, and (c) to provide a thermal neutronflux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, ...

In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutronflux. High thermal neutronfluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

We report on the study of thermal neutronflux using monitors based on mixture of ZnS(Ag) and LiF enriched with a lithium-6 isotope at the deep underground laboratory DULB-4900 at the Baksan Neutrino Observatory. An annual modulation of thermal neutronflux in DULB-4900 is observed. Experimental evidences were obtained of correlation between the long-term thermal neutronflux variations and the absolute humidity of the air in laboratory. The amplitude of the modulation exceed 5\\% of total neutronfluxflux.

Aim: Determination of in-phantom quality factors of Tehran research reactor (TRR) boron neutron capture therapy (BNCT) beam. Materials and Methods: The doses from thermal neutron reactions with 14N and 10B are calculated by kinetic energy released per unit mass approach, after measuring thermal neutronflux using neutron activation technique. Gamma dose is measured using TLD-700 dosimeter. Results: Different dose components have been measured in a head phantom which has been designed an...

Highlights: •Self powered neutron detector (SPND) is attractive neutron monitor for TBM in ITER. •In hard neutron spectra (e.g. TBM) there is the need to optimize their response. •Three state-of-the-art SPNDs were tested using fast and 14 MeV neutrons. •The response of SPNDs is much lower than in thermal neutronflux. •FISPACT calculations performed to find out candidate materials in hard spectra. -- Abstract: Self powered neutron detectors (SPND) have a number of interesting properties (e.g. small dimensions, capability to operate in harsh environments, absence of external bias), so they are attractive neutron monitors for TBM in ITER. However, commercially available SPNDs are optimized for operation in a thermal nuclear reactor where the neutron spectrum is much softer than that expected in a TBM. This fact can limit the use of SPND in a TBM since the effective cross sections for the production of beta emitters are much lower in a fast neutron spectrum. This work represents the first attempt to study SPNDs as neutronflux monitors for TBM. Three state-of-the-art SPND available on the market were bought and tested using fast neutrons at TAPIRO fast neutron source of ENEA Casaccia and with 14 MeV neutrons at the Frascati neutron generator (FNG). The results clearly indicate that in fast neutron spectra, the response of SPNDs is much lower than in thermal neutronflux. Activation calculations were performed using the FISPACT code to find out possible material candidates for SPND suitable for operation in TBM neutron spectra.

In irradiation experiments on materials in the core of the Osiris reactor (CEA-Saclay) we seek to increase damage in irradiated samples and to reduce the duration of their stay in the core. Damage is essentially caused by fast neutrons (E {>=} 1 MeV); we have therefore pursued the possibility of a localized increase of their level in an irradiation experiment by using a flux converter device made up of fissile material arranged according to a suitable geometry that allows the converter to receive experiments. We have studied several parameters that are influential in the increase of fast neutronflux within the converter. We have also considered the problem of the converter's cooling in the core and its effect on the operation of the reactor. We have carried out a specific neutron calculation scheme based on the modular 2D-transport code APOLLO2 using a two-level transport method. Experimental validation of the flux calculation scheme was carried out in the ISIS reactor, the mock-up of OSIRIS, by optimizing the loading of fuel elements in the core. The experimental results show that the neutron calculation scheme computes the fluxes in close agreement with the measurements especially the fast flux. This study allows us to master the essential physical parameters needed for the design of a flux converter in an MTR reactor. (author)

One of the main sources of a background in underground physics experiments (such as the investigation of solar neutrino flux, neutrino oscillations, neutrinoless double beta decay, and the search for annual and daily Cold Dark Matter particle flux modulation) are fast neutrons originating from the surrounding rocks. The measurements of fast neutronflux in the new DULB Laboratory situated at a depth of 4900 m w.e. in the Baksan Neutrino Observatory have been performed. The relative neutron sh...

Full Text Available Americium–beryllium (Am-Be; n, γ is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci, yielding a small thermal neutronflux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutronflux compared to the traditional method with water moderator.

Lead-cooled fast reactor (LFR) has potential for becoming one of the advanced reactor types in the future. Innovative computational tools for system design and safety analysis on such NPP systems are needed. One of the most popular trends is coupling multi-dimensional neutron kinetics (NK) with thermal-hydraulic (T-H) to enhance the capability of simulation of the NPP systems under abnormal conditions or during rare severe accidents. Therefore, various numerical methods applied in the NK module should be reevaluated to adapt the scheme of coupled code system. In the author's present work a neutronic module for the solution of two dimensional steady-state multigroup diffusion problems in nuclear reactor cores is developed. The module can produce both direct fluxes as well as adjoints, i.e. neutron importances. Different numerical schemes are employed. A standard finite-difference (FD) approach is firstly implemented, mainly to serve as a reference for less computationally challenging schemes, such as transverse-integrated nodal methods (TINM) and boundary element methods (BEM), which are considered in the second part of the work. The validation of the methods proposed is carried out by comparisons of the results for some reference structures. In particular a critical problem for a homogeneous reactor for which an analytical solution exists is considered as a benchmark. The computational module is then applied to a fast spectrum system, having physical characteristics similar to the proposed European Lead-cooled System (ELSY) project. The results show the effectiveness of the numerical techniques presented. The flexibility and the possibility to obtain neutron importances allow the use of the module for parametric studies, design assessments and integral parameter evaluations, as well as for future sensitivity and perturbation analyses and as a shape solver for time-dependent procedures

We propose to measure the neutron capture cross sections of $^{139}$La, of $^{93}$Zr (t$_{1/2}$)=1.5 10$^{6}$ yr), and of all the stable Zr isotopes at n_TOF. The aim of these measurements is to improve the accuracy of existing results by at least a factor of three in order to meet the quality required for using the s-process nucleosynthesis as a diagnostic tool for neutron exposure and neutronflux during the He burning stages of stellar evolution. Combining these results with a wealth of recent information coming from high-resolution stellar spectroscopy and from the detailed analysis of presolar dust grains will shed new light on the chemical history of the universe. The investigated cross sections are also needed for technological applications, in particular since $^{93}$Zr is one of the major long-lived fission products.

A prototype compact Liquid Lithium Target (LiLiT), able to constitute an accelerator-based intense neutron source, was built. The neutron source is intended for nuclear astrophysical research, boron neutron capture therapy (BNCT) in hospitals and material studies for fusion reactors. The LiLiT setup is presently being commissioned at Soreq Nuclear research Center (SNRC). The lithium target will produce neutrons through the 7Li(p,n)7Be reaction and it will overcome the major problem of removing the thermal power generated by a high-intensity proton beam, necessary for intense neutronflux for the above applications. The liquid-lithium loop of LiLiT is designed to generate a stable lithium jet at high velocity on a concave supporting wall with free surface toward the incident proton beam (up to 10 kW). During off-line tests, liquid lithium was flown through the loop and generated a stable jet at velocity higher than 5 m/s on the concave supporting wall. The target is now under extensive test program using a high-power electron-gun. Up to 2 kW electron beam was applied on the lithium flow at velocity of 4 m/s without any flow instabilities or excessive evaporation. High-intensity proton beam irradiation will take place at SARAF (Soreq Applied Research Accelerator Facility) superconducting linear accelerator currently in commissioning at SNRC.

In this paper Micromegas has been designed to detect neutrons. The simulation of the spatial reso-lution of Micromegas as neutron detector is carried out by GEANT4 toolkit. The neutron track reconstruction method based on the time coincidence technology is employed in the present work. The influence of the flux of incident 14 MeV neutron and high gamma background on the spatial resolution is carefully studied. Our results show that the spatial resolution of the detector is sensitive to the neutronflux, but insensitive to the intensity of γ background if the neutron track reconstruction method proposed by our group is used. The γ insensitivity makes it possible for us to use the Micromegas detector under condition which has high γ-rays background.

The radial fast neutronflux attenuations in the material (iron) surrounding ducts of diameters 7, 9, and 15 cm and total duct length of about 1.5 m have been investigated with and without neutron scattering cans filled with D{sub 2}O in the duct. Experimentally the problem was solved by the use of foil activation techniques. Theoretically it was attacked by, in the first place, a Monte Carlo program specially written for this purpose and utilizing an importance sampling technique. In the second place non- and single-scattering removal flux codes were tried, and also simple hand calculations. The Monte Carlo results accounted well for the fast flux attenuation, while the non- and single-scattering methods overestimated the attenuation generally by a factor of 10 or less. Simple hand calculations using three empirical parameters could be fitted to the measured data within a factor of 1.2 - 1.3 at penetration depths greater than 3 - 4 cm. The distribution of the D{sub 2}O-scattered flux could well be described in terms of single scattering.

Determination of thermal to fast neutronflux ratio (f(fast)) and fast neutronflux (ϕ(fast)) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f(fast) and subsequently ϕ(fast) were determined using the absolute method. The f(fast) ranged from 48 to 155, and the ϕ(fast) was found in the range 1.03×10(10)-4.89×10(10) n cm(-2) s(-1). These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.

PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutronflux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core with literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.

PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutronflux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core with literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.

The objective is to compare neutron irradiation conditions in Russian reactors and similar US facilities. Neutron fluence and spectral information and calculated radiation damage parameters are presented for the BOR-60 (Fast Experimental Reactor - 60 MW) and SM-2 reactors in Russia. Their neutron exposure characteristics are comparable with those of the Experimental Breeder Reactor (ERB-II), the Fast Flux Test Facility (FFTF), and the High Flux Isotope Reactor (HFIR) in the United States.

Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutronflux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutronflux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutronflux at above mentioned positions. In the present work, the neutronflux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutronfluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

neutron absorption cross section at t. By using this relation in equation (1) the integration is possible over...0 n td f dat) n (it, rpLthprmQJ where 000 is defined as the microscopic neutron absorption cross - section at 2200 m/s, the most probable speed of a... neutron - absorption cross - section of the target as a function of energy O(E) is neutronflux per unit of energy as a function of energy. 1,d is

Measurements of the neutron emission, resulting from nuclear fusion reactions between the hydrogen isotopes deuterium and tritium, can provide a wealth of information on the confinement properties of fusion plasmas and how these are affected by Magneto-Hydro-Dynamic (MHD) instabilities. This thesis describes work aimed to develop neutron measurement techniques for nuclear fusion plasma experiments, specifically regarding the performance and design of collimated neutronflux monitors (neutron ...

We introduce a sequential importance sampling particle filter (PF)-based multisensor multivariate nonlinear estimator for estimating the in-core neutronflux distribution for pressurized heavy water reactor core. Many critical applications such as reactor protection and control rely upon neutronflux information, and thus their reliability is of utmost importance. The point kinetic model based on neutron transport conveniently explains the dynamics of nuclear reactor. The neutronflux in the large core loosely coupled reactor is sensed by multiple sensors measuring point fluxes located at various locations inside the reactor core. The flux values are coupled to each other through diffusion equation. The coupling facilitates redundancy in the information. It is shown that multiple independent data about the localized flux can be fused together to enhance the estimation accuracy to a great extent. We also propose the sensor anomaly handling feature in multisensor PF to maintain the estimation process even when the sensor is faulty or generates data anomaly.

The report describes the new results of the development work performed at Forschungszentrum Karlsruhe on the neutronics of the International Fusion Materials Irradiation Facility (IFMIF). An important step forward has been done in the simulation of neutron production of the deuteron-lithium source using the Li(d,xn) reaction cross sections from evaluated data files. The developed Monte Carlo routine and d-Li reaction data newly evaluated at INPE Obninsk have been verified against available experimental data on the differential neutron yield from deuteron-bombarded thick lithium targets. With the modified neutron source three-dimensional distributions of neutron and photon fluxes, displacement and gas production rates and nuclear heating inside the high flux test module (HFTM) were calculated. In order to estimate the uncertainty resulting from the evaluated data, two independent libraries, recently released by INPE and LANL, have been used in the transport calculations. The proposal to use a reflector around ...

The effect of neutronflux on the formation of irradiation-induced clusters in reactor pressure vessel (RPV) steels is an unresolved issue. Small-angle neutron scattering was measured for a neutron-irradiated RPV weld material containing 0.22 wt% impurity Cu. The experiment was focused on the influence of neutronflux on the formation of irradiation-induced clusters at fixed fluence. The aim was to separate and tentatively interpret the effect of flux on the characteristics of the cluster size distribution. We have observed a pronounced effect of neutronflux on cluster size, whereas the total volume fraction of irradiation-induced clusters is insensitive to the level of flux. The result is compatible with a rate theory model according to which the range of applied fluxes covers the transition from a flux-independent regime at lower fluxes to a regime of decelerating cluster growth. The results are confronted with measured irradiation-induced changes of mechanical properties. Despite the observed flux effect on cluster size, both yield stress increase and transition temperature shift turned out to be independent of flux. This is in agreement with the volume fraction of irradiation-induced clusters being insensitive to the level of flux.

Full Text Available The High flux Advanced Neutron Application ReactOr (HANARO is a unique research reactor in the Republic of Korea, and has been used for irradiation testing since 1998. To conduct irradiation tests for nuclear materials, the irradiation holes of CT and OR5 have been used due to a high fast-neutronflux. Because the neutronflux must be accurately calculated to evaluate the neutron fluence of irradiated material, it was conducted using MCNP. The neutronflux was measured using fluence monitor wires to verify the calculated result. Some evaluations have been conducted, however, more than 20% errors have frequently occurred at the OR irradiation hole, while a good agreement between the calculated and measured data was shown at the CT irradiation hole.

The main purpose of this work is to get an useful guide to carry out topical neutronflux measurements. Although the foil activation technique is used in the majority of the cases, other techniques, such as those based on fission chambers and self-powered neutron detectors, are also shown. Special interest is given to the description and application of corrections on the measurement of relative and absolute induced activities by several types of detectors (scintillators, G-M and gas proportional counters). The thermal arid epithermal neutronfluxes, as determined in this work, are conventional or effective (West cots fluxes), which are extensively used by the reactor experimentalists; however, we also give some expressions where they are related to the integrated neutronfluxes, which are used in neutron calculations. (Author) 16 refs.

Neutron radiography and computed tomography are commonly used techniques to non-destructively examine materials. Tomography refers to the cross-sectional imaging of an object from either transmission or reflection data collected by illuminating the object from many different directions.

The aim of our research is to address why millisecond pulsars have relatively weak surface magnetic fields, of about 10^8 G, with a narrow spread. We propose that the accretion of plasma from the companion star fully screens the original neutron star field, but the accretion disk carries additional magnetic flux from the companion star, or itself can generate field by means of dynamo processes. For a strongly magnetized star, the field prevents the disk from approaching the star. The accretion is along the field lines and deposits the matter on the polar cap. Then, the accreted plasma flows, dragging with itself the magnetic field lines, from the pole to the equator (Payne & Melatos 2004). In a following stage, when the star becomes non-magnetic, because the field has been buried, the disk touches the star. We suggest that some effective mechanism of magnetic flux transport such as that proposed by Spruit & Uzdensky 2005 (or Bisnovatyi-Kogan & Lovelace 2007), operates and necessarily leads to a "strongly magnetized disk''. It becomes laminar because the magneto-rotational instability saturates (it is considered to be responsible for turbulence in the disk), and the magnetic difussivity is negligible. Then, the loss of angular momentum allowing the accretion is only caused by the magneto-centrifugal disk-wind (Blandford & Payne 1982). Meanwhile, the wind-driven transport of the magnetic flux by the disk re-magnetizes the star. This process continues until the Lorentz force due to the star's magnetic field forbids any further accretion of matter and magnetic flux, in the Ideal Magneto-Hydro-Dynamics approach. Additional of material can fall onto the star (but at lower rate) if some instability process sets in, allowing the diffusion of mass through the magnetic field lines (e.g the Interchange Instability, Spruit & Taam 1990). All these processes might lead to an asymptotic magnetic field of 10^8 G,as is inferred from observations. We are developing a self

One of the main sources of a background in underground physics experiments (such as the investigation of solar neutrino flux, neutrino oscillations, neutrinoless double beta decay, and the search for annual and daily Cold Dark Matter particle flux modulation) are fast neutrons originating from the surrounding rocks. The measurements of fast neutronflux in the new DULB Laboratory situated at a depth of 4900 m w.e. in the Baksan Neutrino Observatory have been performed. The relative neutron shielding properties of several commonly available natural materials were investigated too. The preliminary results obtained with a high-sensitive fast neutron spectrometer at the level of sensitivity of about 10^(-7) neutron/ (cm^2 sec) are presented and discussed.

Proton-recoil detectors (PRDs), based on the well known standard H(n,p) elastic scattering cross section, are the preferred instruments to perform precise quasi-absolute neutronflux measurements above 1 MeV. The limitations of using a single silicon detector as PRD at a continuous neutron beam facility are investigated, with the aim of extending such measurements to neutron energies below 1 MeV. This requires a systematic investigation of the background sources affecting the neutronflux measurement. Experiments have been carried out at the AIFIRA facility to identify these sources. A study on the role of the silicon detector thickness on the background is presented and an energy limit on the use of a single silicon detector to achieve a neutronflux precision better than 1% is given.

For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutronflux distribution in the reactor as accurately as possible. The focus of this study is on the neutronflux redistributions due to control rod movements. For analyzing neutronflux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutronflux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system.

Describes an experiment for determining the thermal neutronflux of the light-water nuclear reactor at the University of California, Irvine. The difficulty of the activity can be varied to match the student's level of proficiency. (SL)

The resonance absorption filter technique has been used to determine the thermal/epithermal neutronflux. The main idea in this technique is to use an element with a high and essentially singular resonance in the neutron absorption cross section as a filter surrounding a miniature-type lithium glass scintillator. The count with and without the filter surrounding the detector gives the number of resonance-energy neutrons. Some preliminary results and a comparison with the MCNP code are shown.

Fusion reactors will need high fluxneutron detectors to diagnose the deuterium-deuterium and deuterium-tritium. A candidate detection technique is the Gas Electron Multiplier (GEM). New GEM based detectors are being developed for application to a neutral deuterium beam test facility. The proposed detection system is called Close-contact Neutron Emission Surface Mapping (CNESM). The diagnostic aims at providing the map of the neutron emission due to interaction of the deuterium beam with the deuterons implanted in the beam dump surface. This is done by placing a detector in close contact, right behind the dump. CNESM uses nGEM detectors, i.e. GEM detectors equipped with a cathode that also serves as neutron-proton converter foil. After the realization and test of several small area prototypes, a full size prototype has been realized and tested with laboratory sources. Test on neutron beams are foreseen for the next months.

Fusion reactors will need high fluxneutron detectors to diagnose the deuterium-deuterium and deuterium-tritium. A candidate detection technique is the Gas Electron Multiplier (GEM). New GEM based detectors are being developed for application to a neutral deuterium beam test facility. The proposed detection system is called Close-contact Neutron Emission Surface Mapping (CNESM). The diagnostic aims at providing the map of the neutron emission due to interaction of the deuterium beam with the deuterons implanted in the beam dump surface. This is done by placing a detector in close contact, right behind the dump. CNESM uses nGEM detectors, i.e. GEM detectors equipped with a cathode that also serves as neutron-proton converter foil. After the realization and test of several small area prototypes, a full size prototype has been realized and tested with laboratory sources. Test on neutron beams are foreseen for the next months.

Applying the extreme low-level γ-ray spectroscopic analysis the environmental neutronflux is measured using different moderator construction and environment through the reaction 197Au (n, γ) 198Au. The contribution of thermal and resonance neutrons is separated using the cadmium difference technique, while fast neutrons are measured by the paraffin moderator. The results of altitude dependence of the neutronflux are discussed. The thermal neutronflux near the surface of sea water is less than its value at 100 cm over ground near sea water, while the value over the surfaces of fresh water is higher than that near the surface of sea water. Also the thermal neutronflux at 5 cm soil depth increases, then decreases to its original value at 10 cm depth and still constant until 25 cm, then decreases rapidly to reach 27% of its original value at 60 cm depth. The soil compositions, corresponding neutron temperatures and effective absorption cross sections of earth are the most effective factors on the equilibrium region of thermal neutrons in the ground.

Aim of the project was to find possible explanations for the neutronflux fluctuations and their changes over the last decades in German PWR. Several models concerning thermal hydraulics, structural mechanics and neutron physics were evaluated. It was shown that up to now no models are available that could explain the observed phenomena. Future studies should focus on interdisciplinary coupling of different models.

We revealed an increase in the density of thermal neutronflux in forest biocenoses, which was not associated with astrogeophysical events. The maximum spike of this parameter in the biocenosis reached 10,000 n/(sec x m2). Diurnal pattern of the density of thermal neutronflux depended only on the type of biocenosis. The effects of biomodulation of corpuscular radiation for balneology are discussed.

In this paper, we present calibration methods in order to estimate reactor neutronflux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutronflux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutronflux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

A special neutron diagnostic system is proposed that facilitates the measurement of neutronfluxes and spectra in the neutronics and tritium production-test blanket module (NTTBM) without interrupting the operation of the International Thermal-nuclear Experimental Reactor (ITER),for studying the multiplication rate in the neutron multiplier and breeding ratio of tritium in the breeder.This system includes an encapsulated foil activation system,micro-fission chamber detectors (MFC),and a compact neutron spectrometer using a natural diamond detector (NDD).A helium coolant loop with a reasonable diameter is designed carefully for every measurement channel that ensures that the neutron detectors and preamplifiers would work well under a high temperature scenario and that the filling rates of the neutron multiplier (beryllium pebble)and tritium breeder material (Li4Si04) would not decrease excessively (the expected value≥80%)due to the dimensions of the helium coolant loop.

Nondestructive evaluation methods and systems based on specific attributes of neutron interactions with materials are being developed. The special attributes of neutrons are low attenuation in most engineering materials, strong interaction with low Z elements, and epithermal neutron absorption resonances. The three methods under development at ORNL include neutron based tomography and radiography; through thickness, nondestructive texture mapping; and internal, noninvasive temperature measurement. All three techniques require high flux sources such as the High Flux Isotope Reactor, a steady state source, or the Oak Ridge Electron Linear Accelerator, a pulsed neutron source. Neutrons are quite penetrating in most engineering materials and thus can be useful to detect internal flaws and features. Hydrogen atoms, such as in a hydrocarbon fuel, lubricant, or a metal hydride, are relatively opaque to neutron transmission and thus neutron based tomography/radiography is ideal to image their presence. Texture, the nonrandom orientation of crystalline grains within materials, can be mapped nondestructively using neutron diffraction methods. Epithermal neutron resonance absorption is being studied as a noncontacting temperature sensor. This paper highlights the underlying physics of the methods, progress in development, and the potential benefits for science and industry of the three facilities.

SANS has proved an extremely useful tool for investigating flux-line structures within the bulk of superconductors. With high-T{sub c} materials, the scattered intensities are weak, but careful measurements are giving important new information about flux lattices, flux pinning and flux-lattice melting. (author). 10 refs.

Highlights: ► Neutronic analysis is performed for in situ calibration of the microfission chamber (MFC). ► The source transfer system deigned in this study does not affect MFC detection efficiency. ► The rotation method is appropriate for full calibration because the calibration time is shorter. ► But, point-by-point method should be performed to check the accuracy of the MCNP model. ► Combination of two methods are important to perform in situ calibration efficiently. -- Abstract: Neutronic analysis is performed for in situ calibration of the microfission chamber (MFC), which is the in-vessel neutron-flux monitor at the International Thermonuclear Experimental Reactor (ITER). We present the design of the transfer system for a neutron generator, which consists of two toroidal rings and a neutron-generator holder, and estimate the effect of the system on MFC detection efficiency through neutronic analysis with the Monte Carlo N-particle (MCNP) code. The result indicates that the designed transfer system does not affect MFC detection efficiency. In situ calibrations by the point-by-point method and by the rotation method are also simulated and compared by neutronic analysis. The results indicate that the rotation method is appropriate for full calibration because the calibration time is shorter (all neutron-flux monitors can be calibrated simultaneously). However, the rotation method makes it difficult to compare the results with neutronic analysis, so the point-by-point method should be performed prior to full calibration to check the accuracy of the MCNP model.

This paper describes precise measurements of the thermal neutronflux in the LSM underground laboratory in proximity of the EDELWEISS-II dark matter search experiment together with short measurements at various other locations. Monitoring of the flux of thermal neutrons is accomplished using a mobile detection system with low background proportional counter filled with $^3$He. On average 75 neutrons per day are detected with a background level below 1 count per day (cpd). This provides a unique possibility of a day by day study of variations of the neutron field in a deep underground site. The measured average 4$\\pi$ neutronflux per cm$^{2}$ in the proximity of EDELWEISS-II is $\\Phi_{MB}=3.57\\pm0.05^{stat}\\pm0.27^{syst}\\times 10^{-6}$ neutrons/sec. We report the first experimental observation that the point-to-point thermal neutronflux at LSM varies by more than a factor two.

Benchmark problems were prepared for evaluating the calculation codes and the nuclear data for accelerator shielding design by the Accelerator Shielding Working Group of the Research Committee on Reactor Physics in JAERI. Four benchmark problems: transmission of quasi-monoenergetic neutrons generated by 43 MeV and 68 MeV protons through iron and concrete shields at TIARA of JAERI, neutronfluxes in and around an iron beam stop irradiated by 500 MeV protons at KEK, reaction rate distributions inside a thick concrete shield irradiated by 6.2 GeV protons at LBL, and neutron and hadron fluxes inside an iron beam stop irradiated by 24 GeV protons at CERN are compiled in this document. Calculational configurations and neutron reaction cross section data up to 500 MeV are provided. (author)

The neutronflux of the n{sub T}OF facility at CERN was measured, after installation of the new spallation target, with four different systems based on three neutron-converting reactions, which represent accepted cross sections standards in different energy regions. A careful comparison and combination of the different measurements allowed us to reach an unprecedented accuracy on the energy dependence of the neutronflux in the very wide range (thermal to 1 GeV) that characterizes the n{sub T}OF neutron beam. This is a pre-requisite for the high accuracy of cross section measurements at n{sub T}OF. An unexpected anomaly in the neutron-induced fission cross section of {sup 235}U is observed in the energy region between 10 and 30keV, hinting at a possible overestimation of this important cross section, well above currently assigned uncertainties. (orig.)

Determination of thermal to fast neutronflux ratio (f{sub fast}) and fast neutronflux ({phi}{sub fast}) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f{sub fast} and subsequently {phi}{sub fast} were determined using the absolute method. The f{sub fast} ranged from 48 to 155, and the {phi}{sub fast} was found in the range 1.03x10{sup 10}-4.89x10{sup 10} n cm{sup -2} s{sup -1}. These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.

In this work we show the values distribution of the neutronflux at Peruvian Research Reactor RP-10, determined under two different safety and control rods configurations. The method applied was to irradiate small gold foils in irradiation facilities of the core to carry out the nuclear reaction {sup 197}Au(n, {gamma}){sup 198}Au; then using a gamma spectrometry system and the Westcott formalism we obtained the neutronflux. The results confirm the favorable effect of such configurations, increasing the neutronflux, both thermal and epithermal. These results have consistency with the weekly activity reports of radioisotopes lots given by the Radioisotopes Production Plant and Neutron Activation Analysis Group. (author)

Full Text Available This paper presents some of the challenges ahead of us even after 75 years of the discovery of the fission process and large progress made since then. The focus is on application orientation, which requires improved measurements on fission cross-sections and neutron and γ-ray multiplicities. Experimental possibilities have vastly improved the past decade leading to developments of highly sophisticated detector systems and the use of digital data acquisition and signal processing. The development of innovative fast nuclear reactor technology needs improved respective nuclear data. Advancements in theoretical modelling also require better experimental data. Theory has made progress in calculating fission fragment distributions (i.e. GEF code as well as prompt neutron and γ-ray emission to catch up with the improved experiments.

To achieve wide-range, high-integration, and real-time performance on the neutronflux measurement on the HL-2A tokamak, a digital neutronflux measuring (DNFM) system based on the peripheral component interconnection (PCI) eXtension for Instrumentation express (PXIe) bus was designed. This system comprises a charge-sensitive preamplifier and a field programmable gate array (FPGA)-based main electronics plug-in. The DNFM totally covers source-range and intermediate-range neutronflux measurements, and increases system integration by a large margin through joining the pulse-counting mode and Campbell mode. Meanwhile, the neutronflux estimation method based on pulse piling proportions is able to choose and switch measuring modes in accordance with current flux, and this ensures the accuracy of measurements when the neutronflux changes suddenly. It has been demonstrated by simulated signals that the DNFM enhances the full-scale measuring range up to 1.9 × 108 cm-2 s-1, with relative error below 6.1%. The DNFM has been verified to provide a high temporal sensitivity at 10 ms time intervals on a single fission chamber on HL-2A. Contributed paper, published as part of the Proceedings of the 3rd Domestic Electromagnetic Plasma Diagnostics Workshop, September 2016, Hefei, China.

Neutron measurement results obtained at SINP MSU since 1970 are presented. These measurements were made using techniques based on neutron moderation and subsequent detection in a Li6I(Eu) crystal or a He3 coronal counter. The measurements were mainly carried out in orbits with inclination of 52 degrees and altitudes of 200-450 km. The spatial and angular distributions of the measured neutronfluxes were studied. The albedo neutronflux was estimated according to the count rate difference for opposite detector orientations towards Earth and away from it. This flux is comparable to the local neutronflux outside the Brazil anomaly region, where local neutrons dominate. Neutronfluxes, generated by solar protons, were detected during a solar flare on June 6, 1991 for the first time. Their spectrum was estimated as a power law with alpha>2.

A sophisticated neutron guide system has been installed at the new Munich neutron source FRM-II to transport neutrons from the D 2 cold neutron source to several instruments, which are situated in a separate neutron guide hall. The guide system takes advantage of supermirror coatings and includes a worldwide unique "twisted" guide for a desired phase space transformation of the neutron beam. During the initial reactor commissioning in summer 2004, the integral and differential neutronflux as well as the distribution of beam divergence at the exit of two representative and the twisted neutron guide were measured using time-of-flight spectroscopy and gold-foil activation. The experimental results can be compared to extensive simulation calculations based on MCNP and McStas. The investigated guides fulfill the expectations of providing high neutronfluxes and reveal good quality with respect to the reflective coatings and the installation precision.

A sophisticated neutron guide system has been installed at the new Munich neutron source FRM-II to transport neutrons from the D{sub 2} cold neutron source to several instruments, which are situated in a separate neutron guide hall. The guide system takes advantage of supermirror coatings and includes a worldwide unique 'twisted' guide for a desired phase space transformation of the neutron beam. During the initial reactor commissioning in summer 2004, the integral and differential neutronflux as well as the distribution of beam divergence at the exit of two representative and the twisted neutron guide were measured using time-of-flight spectroscopy and gold-foil activation. The experimental results can be compared to extensive simulation calculations based on MCNP and McStas. The investigated guides fulfill the expectations of providing high neutronfluxes and reveal good quality with respect to the reflective coatings and the installation precision.

Two types of neutron monitors with fine spatial resolution are proposed based on vibrating wire. In the first type, neutrons interact with the vibrating wire, heat it, and lead to the change of natural frequency, which can be precisely measured. To increase the heat deposition during the neutron scattering, use of gadolinium layer which has the highest thermal neutron capture cross section among all elements is proposed. The second type of the monitor uses vibrating wire as a resonant target. Besides the measurement of beam profile according to the average signal, the differential signal synchronized with the wire oscillations defines the gradient of beam profile. Spatial resolution of the monitor is defined by the diameter of the wire.

We propose to provide theoretical support for the experimental program in fundamental neutron physics at the SNS. This includes the study of neutron properties, neutron beta-decay, parity violation effects and time reversal violation effects. The main purpose of the proposed research is to work on theoretical problems related to experiments which have a high priority at the SNS. Therefore, we will make a complete analysis of beta-decay process including calculations of radiative corrections and recoil corrections for angular correlations for polarized neutron decay, with an accuracy better that is supposed to be achieved in the planning experiments. Based on the results of the calculations, we will provide analysis of sensitivity of angular correlations to be able to search for the possible extensions of the Standard model. Also we will help to plan other experiments to address significant problems of modern physics and will work on their theoretical support.

In this paper we analyze the neutron emission observations made in the experiment onboard MIR orbital station (1991), ISS (2002) and Colibri-2002 satellite (2002) at the altitude of 400 km. The helium discharge detectors made it possible to detect neutrons with energies ranging from 0.25eV to 1.9MeV. The spatial distribution of high background neutronfluxes has a longitude dependence. These events have been observed at -200 ... 600 and 1350 ...1800 ...- 1350 longitudinal intervals. The most intensive fluxes near the geomagnetic equator were registered in the African region. They are not found to be associated with increases of proton fluxes (Ep >50MeV). As a statistical set, the events appear to coincide with the most active region of atmospheric weather. In this paper we assess the possibility that the occurrence of high background neutronfluxes in the African region is connected with lightning discharges. To observe neutron emission at the altitude of 400 km ~101 0 neutrons are required to be produced by lightning discharge. These theoretical predictions suggest cloud charge values of about 250-300 Coulomb.

The Avalanche Photodiodes ( APD) were chosen as photon sensors for the region of the CMS electromagnetic calorimeter. The LHC will be a hard environment for what concerns the radiation levels in the detectors. The most relevant damage on APDs is caused by neutrons that produce an increase in the dark current of these devices. In the CMS-ECAL collaboration a big effort was indeed done to understand this damage, but the evaluation of the absolute effect was limited by the knowledge of the neutronflux calibration of the various irradiation facilities. This investigation describes the calibration of the neutronflux of the Tapiro reactor in Rome and the calculation of the Non-Ionizing-Energy-Loss on Silicon for this reactor. The damage parameter alpha for the APDs is evaluated to be about 10-11*10^-17 A/cm/neutron at 18C and 2 days after the irradiation. Some cross-checks with other irradiation facilities are also presented.

The 3rd irradiation test of DUPIC fuel, which was fabricated in the DFDF(DUPIC Fuel Development Facility) was performed in HANARO. For the objectives of this irradiation test, the newly designed irradiation rig was equipped with three Rh- type SPND sensors around DUPIC mini-elements for estimating the thermal neutronflux in the OR4 hole. The thermal neutronflux was measured at this location for 5 months the start of the test. The measured data were transmitted to monitoring system. We confirmed that the trend of SPND signal is well agree with that of HANARO power. The measured average thermal neutronflux is 0.45 n/cm{sup 2} {center_dot}s and the average linear power of DUPIC mini-element was estimated to be 33.5 KW/m.

The term `thermal flux' implies a Maxwellian distribution of velocity and energy corresponding to the most probable velocity of 2200 ms-1 at 293.4 K. In order to measure the thermal neutronflux density, the foil activation method was used. Thermal neutronflux determination in paraffin phantom by counting the emitted rays of indium foils with two different detectors (Geiger–Muller counter and NaI(Tl)) was the aim of this project. The relative differences of the outcome of the experiments were between 2.5% and 5%. The final results were compared with MCNP4C outputs and the best agreement was generated using NaI(Tl) by a minimum discrepancy of about 0.6% for the foil placed 8.5 cm from the neutron source.

The variation of remotely sensed neutron count rates is measured as a function of cratercentric distance using data from the Lunar Prospector Neutron Spectrometer. The count rate, stacked over many craters, peaks over the crater centre, has a minimum near the crater rim and at larger distances it increases to a mean value that is up to 1% lower than the mean count rate observed over the crater. A simple model is presented, based upon an analytical topographical profile for the stacked craters fitted to data from the Lunar Orbiter Laser Altimeter (LOLA). The effect of topography coupled with neutron beaming from the surface largely reproduces the observed count rate profiles. However, a model that better fits the observations can be found by including the additional freedom to increase the neutron emissivity of the crater area by ~0.35% relative to the unperturbed surface. It is unclear what might give rise to this effect, but it may relate to additional surface roughness in the vicinities of craters. The ampl...

Full Text Available Some exotic geophysical events are observed by a global net of electron-neutron detectors (en-detectors developed in the framework of the PRISMA EAS project. Our en-detectors running both on the Earth's surface and underground are continuously measuring the environmental thermal neutronflux. Thermal neutrons are in equilibrium with media and are therefore sensitive to many geophysical phenomena, which are exotic for people studying ultra high-energy cosmic rays or carrying out low background experiments deep underground.

A semi-Lagrangian method for parabolic problems is proposed, that extends previous work by the authors to achieve a fully conservative, flux-form discretization of linear and nonlinear diffusion equations. A basic consistency and convergence analysis are proposed. Numerical examples validate the proposed method and display its potential for consistent semi-Lagrangian discretization of advection--diffusion and nonlinear parabolic problems.

The Joint Research Centre has taken into operation a new experimental device designed for research in the fields of nuclear safeguards and security applications. The research projects currently undertaken include detection of shielded contraband materials, detection of fissile materials, and mass determination of small fissile materials in shielded containers. The device, called the Pulsed Neutron Interrogation Test Assembly (PUNITA), incorporates a pulsed 14-MeV (D-T) neutron generator and a large graphite mantle surrounding the sample cavity. By pulsing the neutron generator with a frequency in the range of 10 to 150 Hz, a sample may be interrogated first by fast neutrons and a few hundred micro-seconds later by a pure thermal neutronflux. The permanent detection systems incorporated in PUNITA include 3He neutrons detectors, HPGe gamma detectors, and lanthanum bromide scintillation detectors. We have studied the effects of placing a tungsten liner around the neutron generator target. The 14-MeV neutrons induce (n, 2n) and (n, 3n) reactions. In addition the mean neutron energy emitted from generator/tungsten assembly is reduced to about 1 MeV. Both of these effects increase the thermal neutronflux in the sample cavity. The paper describes the observed advantages of the tungsten liner with respect to increase in thermal flux, and better shielding capabilities of the nearby gamma and neutron detectors.

Cancer treatment is performed, in Particle Therapy, using accelerated charged particles whose high irradiation precision and conformity allows the tumor destruction while sparing the surrounding healthy tissues. Dose release monitoring devices using photons and charged particles produced by the beam interaction with the patient body have already been proposed, but no attempt based on the detection of the abundant secondary radiation neutron component has been made yet. The reduced attenuation length of neutrons yields a secondary particle sample that is larger in number when compared to photons and charged particles. Furthermore, neutrons allow for a backtracking of the emission point that is not affected by multiple scattering. Since neutrons can release a significant dose far away from the tumor region, a precise measurement of their flux, production energy and angle distributions is eagerly needed in order to improve the Treatment Planning Systems (TPS) software, so to predict not only the normal tissue toxicity in the target region but also the risk of late complications in the whole body. All the aforementioned issues underline the importance for an experimental effort devoted to the precise characterization of the neutron production gaining experimental access both to the emission point and production energy. The technical challenges posed by a neutron detector aiming for a high detection efficiency and good backtracking precision will be addressed within the MONDO (MOnitor for Neutron Dose in hadrOntherapy) project. The MONDO's main goal is to develop a tracking detector targeting fast and ultrafast secondary neutrons. The tracker is composed by a scintillating fiber matrix (4 × 4 × 8cm3). The full reconstruction of protons, produced in elastic interactions, will be used to measure energy and direction of the impinging neutron. The neutron tracker will measure the neutron production yields, as a function of production angle and energy, using different

A proof-of-principle collimated Neutronflux Camera (NC) monitor for the measurement of the 2.45 MeV neutron emission from the deuterium-deuterium fusion reactions has been developed, installed and put into use at the Mega Ampere Spherical Tokamak (MAST). The NC measures the spatial and time resolved volume integrated neutron emissivity in deuterium fusion plasmas in the presence of auxiliary plasma heating along two equatorial and two diagonal lines of sight whose tangency radius can be varied between plasma discharges. This paper describes the NC design principles, their technical realization and its performances illustrated with experimental observations of different plasma scenarios. Neutron count rates in the range 0.1-1.5 MHz are routinely observed allowing time resolutions as high as 1 ms with a statistical uncertainty less than 10% and an energy threshold of 0.5 MeV. Examples of the effect of plasma instabilities on the neutron emission are presented. The good results obtained will be used for the design of the neutronflux camera monitor for MAST Upgrade.

A proof-of-principle collimated Neutronflux Camera (NC) monitor for the measurement of the 2.45 MeV neutron emission from the deuterium–deuterium fusion reactions has been developed, installed and put into use at the Mega Ampere Spherical Tokamak (MAST). The NC measures the spatial and time resolved volume integrated neutron emissivity in deuterium fusion plasmas in the presence of auxiliary plasma heating along two equatorial and two diagonal lines of sight whose tangency radius can be varied between plasma discharges. This paper describes the NC design principles, their technical realization and its performances illustrated with experimental observations of different plasma scenarios. Neutron count rates in the range 0.1–1.5 MHz are routinely observed allowing time resolutions as high as 1 ms with a statistical uncertainty less than 10% and an energy threshold of 0.5 MeV. Examples of the effect of plasma instabilities on the neutron emission are presented. The good results obtained will be used for the design of the neutronflux camera monitor for MAST Upgrade.

Computer program contains an input library of nuclear data for 44 elements and their isotopes to determine the induced radioactivity for gamma emitters. Minimum input requires the irradiation history of the element, a four-energy-group neutronflux, specification of an alloy composition by elements, and selection of the output.

We describe the use of reverse Monte Carlo refinement to extract structural information from angle-resolved data of a Bragg peak. Starting with small-angle neutron scattering data, the positional order of an ensemble of flux lines in superconducting Nb is revealed. We discuss the uncovered correl...

Methods used to calculate the effects of exposure to a neutronflux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated.

A charge-collection method for measuring the flux of pulsed fast neutrons in current mode has been developed, which is based on the well-known recoil-proton method combined with ion-induced secondary electron emission from solid surfaces. The detection unit consists of four elements: an n-p converter, an absorber, a collector, and a rear insulator. The assembly does not require vacuum for operation. Recoil protons from the n-p converter and the secondary electrons induced by the passing protons on the interface of the absorber and the collector contribute to the detector output signal. By properly choosing the materials and the combination of the absorber and the collector, the fraction of secondary electrons in the output signal can be determined experimentally. This detection concept allows one to design a medium type of fast-neutron detector for measurements of extremely intense pulsed neutronflux with a number of advantages over the existing systems.

Air radon concentration measurement is useful for research on geophysical effects, but it is strongly sensitive to site geology and many geophysical and microclimatic processes such as wind, ventilation, air humidity and so on that induce very big fluctuations on the concentration of radon in air. On the contrary, monitoring the radon concentration in soil by measuring the thermal neutronflux reduces environmental effects. In this paper we report some experimental results on the natural thermal neutronflux as well as the concentration of air radon and its variations at 4300 m a.s.l. These results were obtained with unshielded thermal neutron scintillation detectors (en-detectors) and radon monitors located inside the ARGO-YBJ experimental hall. The correlation of these variations with the lunar month and 1-year period is undoubtedly confirmed. A method for earthquakes prediction provided by a global net of the en-detectors is currently under study.

Americium–beryllium (Am-Be; n, γ) is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences...

Aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann Equation governing the Laplace transform (with respect to the lethargy variable) of the neutronflux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. Use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u/sup -5/. The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, represent one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M/sub 2/(u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table.

Next-generation spallation neutron source facilities will offer instruments with unprecedented capabilities through simultaneous enhancement of source power and usage of advanced optical components. The Spallation Neutron Source (SNS), already under construction at Oak Ridge National Laboratory and scheduled to be completed by 2006, will provide greater than an order of magnitude more effective source flux than current state-of-the-art facilities, including the most advanced research reactors. An additional order of magnitude gain is expected through the use of new optical devices and instrumentation concepts. Many instrument designs require supermirror neutron guides with very high critical angles for total reflection. In this contribution, we will discuss how the performance of a modern neutron-scattering instrument depends on the efficiency of these supermirrors. We summarize current limitations of supermirror coatings and outline ideas for enhancing their performance, particularly for improving the reflec...

A pulse shape discrimination method to discriminate neutron events from backgrounds based on the double-pulse effect of 3He proportional counters is proposed and detailed in this paper. We made an ambient thermal neutron measurement system composed of a commercial 3He proportional counter tube and the corresponding readout electronics. The background of the system has been measured and the minimum detectable amount of the 3He proportional counter tube will be reduced by an order of magnitude with this method. The system was applied to measure the ambient thermal neutronflux inside a large neutron shielding structure at a deep underground laboratory and the pulse shape discrimination method proves to be effective.

Polarized {sup 3}He has shown its unique characteristics in many areas of polarized neutron scattering, its ability to polarize neutrons at short wavelengths, accept wide-angle and divergent beams and low backgrounds enable new classes of experiments. While polarized {sup 3}He is not a steady state solution as commonly applied, the benefits have been shown to offset the drawbacks of polarizing and refreshing the polarization in the neutron spin filter cells. As an extension of this work, in-situ polarization using the spin-exchange optical pumping (SEOP) method was explored as a means to construct a system which could be used to polarize {sup 3}He in the state used for an effective neutron spin filter to constant polarization while on the neutron beam. An in-situ SEOP polarizer was constructed. This device utilized many devices and principles developed for neutron spin filters which are polarized off the beam line using either SEOP or metastability exchange optical pumping (MEOP) under the same research program. As a collimation of this work effects of extremely high neutron capture flux density >1x10{sup 10}cm{sup -2}s{sup -1} incident on the in-situ polarizer were explored.

Although the fission of heavy nuclei was discovered over 75 years ago, many problems and questions still remain to be addressed and answered. The reader will be presented with an old, but persistent problem of this field: The contradiction between Prompt Fission Neutron (PFN) spectra measured with differential (microscopic) experiments and integral (macroscopic and benchmark) experiments (the Micro-Macro problem). The difference in average energy is rather small ~3% but it is stable and we cannot explain the difference due to experimental uncertainties. Can we measure the PFN spectrum with hig

Full Text Available The present work contains some results of observations of neutronflux variations near the Earth’s surface. The Earth’s crust is determined to be a significant source of thermal and slow neutrons, originated from the interaction between the nuclei of the elements of the Earth’s crust and the atmosphere and α-particles, produced by decay of radioactive gases (Radon, Thoron and Actinon. In turn, variations of radioactive gases exhalation is connected with geodynamical processes in the Earth’s crust, including tectonic activity. This determined relation between the processes in the Earth’s crust and neutrons’ flux allow to use variations of thermal and slow neutrons’ flux in order to observe increasing tectonic activity and to develop methods for short-term prediction of natural hazards.

In JMTR (Japan Materials Testing Reactor, 50 MW), the core configuration has been changed from previously employed Mixed-core (25 LEUs(low enrichment uranium (19.8%) fuel elements) and 2 MEUs (medium enrichment uranium (45%) fuel elements)) to MEU6-core (21 LEUs and 6 MEUs), since 125th operating cycle (started in Nov. 17, 1998). In order to investigate the effect of core configuration change on the irradiation tests, neutronflux distribution and neutron spectrum of irradiation field in MEU6-core were calculated by diffusion code CITATION and Monte Carlo code MCNP. As the result, it was confirmed that irradiation field in the MEU6-core has the neutronic characteristics almost equivalent to the irradiation field in the Mixed-core. (author)

Full Text Available In cancer treatment, high energy X-rays are used which are produced by linear accelerators (LINACs. If the energy of these beams is over 8 MeV, photonuclear reactions occur between the bremsstrahlung photons and the metallic parts of the LINAC. As a result of these interactions, neutrons are also produced as secondary radiation products (γ,n which are called photoneutrons. The study aims to map the photoneutron flux distribution within the LINAC bunker via neutron activation analysis (NAA using indium-cadmium foils. Irradiations made at different gantry angles (0°, 90°, 180° and 270° with a total of 91 positions in the Philips SLI-25 linear accelerator treatment room and location-based distribution of thermal neutronflux was obtained. Gamma spectrum analysis was carried out with high purity germanium (HPGe detector. Results of the analysis showed that the maximum neutronflux in the room occurred at just above of the LINAC head (1.2x105 neutrons/cm2.s which is compatible with an americium-beryllium (Am-Be neutron source. There was a 90% decrease of flux at the walls and at the start of the maze with respect to the maximum neutronflux. And, just in front of the LINAC door, inside the room, neutronflux was measured less than 1% of the maximum.

In this paper, a new high transparency device based on MCP for the monitoring the flux and spatial profile of a neutron beam will be described. The assembly consists of a carbon foil with a 6Li deposit, placed in the beam, and a MCP equipped with a phosphor screen readout viewed by a CCD camera, placed outside the beam. Secondary emitted electrons (SEE) produced in the carbon foil by the alpha-particles and tritons from the 6Li+n reaction, are deflected to the MCP detector by means of an electrostatic mirror, suitably designed to preserve the spatial resolution. The conductive layer on the phosphor can be used for neutron counting, and to obtain time-of-flight information. A peculiar feature of this device is that the use of an electrostatic mirror minimizes the perturbation of the neutron beam, i.e. absorption and scattering. It can be used at existing time-of-flight (TOF) facilities, in particular at the n_TOF facility at CERN, for monitoring the flux and special profile of the neutron beam in the thermal and epithermal region. In this work, the device principle and design will be presented, together with the main features in terms of resolution and neutron detection efficiency.

Low energy accelerators in underground locations have emerged as a powerful tool for the measurement of critical nuclear reactions for the study of energy production and element synthesis in astrophysics. While cosmic ray induced background is substantially reduced, beam induced background on target impurities and depositions on target and collimator materials remain a matter of serious concern. The Dual Ion Accelerator for Nuclear Astrophysics (DIANA) is proposed to operate as a low-level background facility in an underground location. One of the main goals of DIANA is the study of neutron sources in stellar helium burning. For these experiments DIANA is a neutron radiation source which may affect other nearby low background level experiments. We therefore investigated the required laboratory layout to attenuate the neutronflux generated in a worst-case scenario to a level below the natural background in the underground environment. Detailed Monte Carlo calculations of the neutron propagation in the laboratory show that a neutronflux many orders of magnitude above expected values gets attenuated below the natural background rate using a 1 m thick water-shielded door as well as an emergency access/egress maze.

Full Text Available A semi-Lagrangian method for parabolic problems is proposed, that extends previous work by the authors to achieve a fully conservative, flux-form discretization of linear and nonlinear diffusion equations. A basic consistency and stability analysis is proposed. Numerical examples validate the proposed method and display its potential for consistent semi-Lagrangian discretization of advection diffusion and nonlinear parabolic problems.

Full Text Available The H5B is a concept of an accelerator-driven sub-critical research facility (ADSRF being developed over the last couple of years at the Vinča Institute of Nuclear Sciences, Belgrade, Serbia. Using well-known computer codes, the MCNPX and MCNP, this paper deals with the results of a tar get study and neutronflux calculations in the sub-critical core. The neutron source is generated by an interaction of a proton or deuteron beam with the target placed inside the sub-critical core. The results of the total neutronflux density escaping the target and calculations of neutron yields for different target materials are also given here. Neutrons escaping the target volume with the group spectra (first step are used to specify a neutron source for further numerical simulations of the neutronflux density in the sub-critical core (second step. The results of the calculations of the neutron effective multiplication factor keff and neutron generation time L for the ADSRF model have also been presented. Neutron spectra calculations for an ADSRF with an uranium tar get (highest values of the neutron yield for the selected sub-critical core cells for both beams have also been presented in this paper.

A fast time response, wide dynamic range neutronflux monitor has been developed toward the LHD deuterium operation by using leading-edge signal processing technologies providing maximum counting rate up to ∼5 × 10{sup 9} counts/s. Because a maximum total neutron emission rate over 1 × 10{sup 16} n/s is predicted in neutral beam-heated LHD plasmas, fast response and wide dynamic range capabilities of the system are essential. Preliminary tests have demonstrated successful performance as a wide dynamic range monitor along the design.

The SciBath detector is an 80 liter liquid scintillator detector read out by a three dimensional grid of 768 wavelength-shifting fibers. Initially conceived as a fine-grained charged particle detector for neutrino studies that could image charged particle tracks in all directions, it is also sensitive to fast neutrons (15-200 MeV). In fall of 2011 the apparatus performed a three month run to measure cosmic-induced muons and neutrons 100~meters underground in the FNAL MINOS near-detector area. Data from this run has been analyzed and resulted in measurements of the cosmic muon flux as \

Micro-pocket fission detectors (MPFD) have been fabricated and tested as in-core flux monitors in the 250 kW TRIGA nuclear reactor at Kansas State University. The prototype devices have been coated with a natural uranyl-nitrate to provide a neutron reactive coating. The devices are composed of alumina substrates sealed together to form a miniature gas pocket 3 mm in diameter and 1 mm wide. The devices are radiation hard and can operate in pulse mode in a neutronflux exceeding 10{sup 12} cm{sup -2} s{sup -1}. Placed in the central thimble of the reactor core, the MPFDs have shown count rate linearity from low to high power. Dead time losses become apparent at power levels exceeding 100 kW, yet are still low enough to allow for pulse mode operation.

There are some ideas about the Earth’s global seismic activity appearance due to tidal forces. At the same time, the correlations between the big series of the earthquakes and the New and Full Moons and between the New and Full Moons and the increasings of the thermal neutrons’ flux from the Earth’s crust were observed. It is as though there are internal links between these three natural phenomena and the physical reasons for their appearance are the same. The paper presents the results of the ground-based thermal neutrons observations during different time periods characterized with phenomena in the near-Earth space (for instance, the New and Full Moon). Basing on the up-to-date conception of the tidal waves influence on the Earth's crust the authors confirm the role of the Moon in the production of the neutronflux near the Earth's surface.

The High Flux Isotope Reactor resumed operation in June of 2007 with a supercritical hydrogen cold source in horizontal beam tube 4. Cold guide 4 is a guide system designed to deliver neutrons from this source with a reasonable flux at wavelengths greater than 4 A to several instruments, and includes a 15-m, 96-section, 4-channel bender. A time-of-flight spectrum with calibrated detector was recorded at port C of cold guide 4, and compared to McStas simulations, to generate a brightness spectrum.

The High Flux Isotope Reactor resumed operation in June of 2007 with a super-critical hydrogen cold source in horizontal beam tube 4. Cold guide 4 is a guide system designed to deliver neutrons from this source at reasonable flux at wavelengths greater than 4 Å to several instruments, and includes a 15-m, 96-section, 4-channel bender. A time-of-flight spectrum with calibrated detector was recorded at port C of cold guide 4, and compared to McStas simulations, to generate a brightness spectrum.

The High Flux Isotope Reactor resumed operation in June of 2007 with a supercritical hydrogen cold source in horizontal beam tube 4. Cold guide 4 is a guide system designed to deliver neutrons from this source with a reasonable flux at wavelengths greater than 4 Å to several instruments, and includes a 15-m, 96-section, 4-channel bender. A time-of-flight spectrum with calibrated detector was recorded at port C of cold guide 4, and compared to McStas simulations, to generate a brightness spectrum.

The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutronfluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutronflux and the epithermal index in each site. The calculated and measured results agree well.

Radiation hardness is an important requirement for solid state readout devices operating in high radiation environments common in particle physics experiments. The MEG II experiment, at PSI, Switzerland, investigates the forbidden decay μ+ → e+ γ. Exploiting the most intense muon beam of the world. A significant flux of non-thermal neutrons (kinetic energy Ek>= 0.5 MeV) is present in the experimental hall produced along the beam-line and in the hall itself. We present the effects of neutronfluxes comparable to the MEG II expected doses on several Silicon Photomultiplier (SiPMs). The tested models are: AdvanSiD ASD-NUV3S-P50 (used in MEG II experiment), AdvanSiD ASD-NUV3S-P40, AdvanSiD ASD-RGB3S-P40, Hamamatsu and Excelitas C30742-33-050-X. The neutron source is the thermal Sub-critical Multiplication complex (SM1) moderated with water, located at the University of Pavia (Italy). We report the change of SiPMs most important electric parameters: dark current, dark pulse frequency, gain, direct bias resistance, as a function of the integrated neutron fluency.

A neutron radiography and tomography facilityhave been developed recently at the IBR-2 high flux pulsed reactor. The facility is operated with the CCD-camera based detector having maximal field of view of 20x20 cm, and the L/D ratio can be varied in the range 200 - 2000. The first results of the radiography and tomography experiments with industrial materials and products, paleontological and geophysical objects, meteorites, are presented.

A method to determine the flux and angular distribution of thermal neutrons with the use of Lr-115 detectors was developed. The use of the Lr-115 detector involves the exposure of a pressed boric acid sample (tablet) as a target, in tight contact with the track detector, to a flux of thermalized neutrons. The self-absorption effects in thin films or foil type thermal neutron detectors can be neglected by using the Lr-115 detector and boric acid tablet setup to operate via backside irradiation. The energy window and the critical angle-residual energy curve were determined by comparisons between the experimental and simulated track parameters. A computer program was developed to calculate the detector registration efficiency, so that the thermal neutronflux can be calculated from the track densities induced in the Lr-115 detector using the derived empirical formula. The proposed setup can serves as directional detector of thermal neutrons. (Author)

Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutronfluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one

During 1988 the first positive steps were taken to proceed with the design and construction of a neutron capture therapy facility on the High Flux Reactor (HFR) at Petten. The immediate aim is to realise within a short time (summer 1989), an epithermal neutron beam for radiobiological and filter optimisation studies on one of the 10 small aperture horizontal beam tubes. The following summer, a much larger neutron beam, i.e., in cross section and neutron fluence rate, will be constructed on one of the two large beam tubes that replaced the old thermal column in 1984. This latter beam tube faces one whole side of the reactor vessel, extending from a 50 x 40 cm input aperture to a 35 x 35 cm exit hole. The radiotherapeutic facility will be housed here, with the intention to start clinical trials at the beginning of 1991. This paper describes the present status of the project and includes: a general description of the pertinent characteristics with respect to NCT of the HFR; results of the recently completed preliminary neutron metrology and computer modeling at the input end of the candidate beam tube; the structure and planning of the proposed Work Programme; and the respective direct and indirect participation and collaboration with the Netherlands Cancer Institute and the European Collaboration Group on BNCT.

Highlights: > The photo-neutron source was investigated within Ghana MNSR irradiation channels. > Irradiation channels under study were inner, outer and the fission chamber. > Thermal rated power at sub-critical state was estimated. > Neutronflux variation was investigated within the channels. > MCNP code has been used to investigate the flux variation. - Abstract: Computer simulation was carried out for photo-neutron source variation in outer irradiation channel, inner irradiation channels and the fission channel of a tank-in-pool reactor, a Miniature Neutron Source Reactor (MNSR) in sub-critical condition. Evaluation of the photo-neutron was done after the reactor has been in sub-critical condition for three month period using Monte Carlo Neutron Particle (MCNP) code. Neutronflux monitoring from the Micro Computer Control Loop System (MCCLS) was also investigated at sub-critical condition. The recorded neutronfluxes from the MCCLS after investigations were used to calculate the power of the reactor at sub-critical state. The computed power at sub-critical state was used to normalize the un-normalized results from the MCNP.

In this paper, we study the monitoring method of neutronflux for the spallation target used in an accelerator driven sub-critical (ADS) system, where the spallation target located vertically at the centre of a sub-critical core is bombarded vertically by the high-energy protons from an accelerator. First, by considering the characteristics in the spatial variation of neutronflux from the spallation target, we propose the following multi-point measurement technique, i.e. the spallation neutronflux should be measured at multiple vertical locations. To explain why the flux should be measured at multiple locations, we have studied the neutron production from tungsten target bombarded by a 250 MeV-proton beam with the Geant4-based Monte Carlo simulations. The simulation results have indicated that the neutronflux at the central location is up to three orders of magnitude higher than the flux at the lower locations. Secondly, we have developed an effective technique in order to measure the spallation neutron fl...

Full Text Available Neutron activation analysis is widely used for identification of elements and their quantities even in trace amounts in the samples of almost any type. The challenges in detecting trace amounts of particular elements are often associated with the neutronflux produced at the research reactors. Low-fluxneutron activation analysis usually presents the biggest challenge when analyzing trace quantities of elements with lower magnitude of radiative capture cross-sections. In this paper, we present the methodology and the quantified uncertainties associated with the detection of trace amounts of cobalt and europium, using as an example concrete aggregates. Recent growing interest is in improving structural concrete (increasing its strength but reducing its activation in nuclear power plant environments. Aside from buildings, structural concrete is also used as a biological shield in nuclear power plant that become radioactive after exposure to neutronflux. Due to radiative capture interactions, artificial radionuclides are generated to high enough concentrations that classify concrete as low-level radioactive waste at the time of the plant's decommissioning. Disposal of this concrete adds to the expense of nuclear power plant financing and its construction. Three radionuclides, 60Co, 152Eu, and 154Eu, account for 99 % of total residual radioactivity of nuclear power plant decommissioned concrete. IAEA document RS-G-1.7, Application of the Concepts of Exclusion, Exemption, and Clearance, specifies clearance levels of radionuclides specific activities: a specific activity lower than 0.1 Bqg-1 for 60Co and 152Eu, and 154Eu allows for a concrete to be recycled after decommissioning of the nuclear power plant. Therefore, low-fluxneutron activation analysis is used to test the detection limits of trace elements in samples of cement, coarse, and fine concrete aggregates. These samples are irradiated at the University of Utah's 100 kW TRIGA Reactor at

Calculations of radiation flux and dose distributions for Boron Neutron Capture Therapy (BNCT) of brain tumors are typically performed using sophisticated three-dimensional analytical models based on either a homogeneous approximation or a simplified few-region approximation to the actual highly-heterogeneous geometry of the irradiation volume. Such models should be validated by comparison with calculations using detailed models in which all significant macroscopic tissue heterogeneities and geometric structures are explicitly represented as faithfully as possible. This work describes a validation exercise for BNCT of canine brain tumors. Geometric measurements of the canine anatomical structures of interest for this work were performed by dissecting and examining two essentially identical Labrador Retriever heads. Chemical analyses of various tissue samples taken during the dissections were conducted to obtain measurements of elemental compositions for tissues of interest. The resulting geometry and tissue composition data were then used to construct a detailed heterogeneous calculational model of the Labrador Retriever head. Calculations of three-dimensional radiation flux distributions pertinent to BNCT were performed for the model using the TORT discrete-ordinates radiation transport code. The calculations were repeated for a corresponding volume-weighted homogeneous tissue model. Comparison of the results showed that the peak neutron and photon flux magnitudes were quite similar for the two models (within 5%), but that the spatial flux profiles were shifted in the heterogeneous model such that the fluxes in some locations away from the peak differed from the corresponding fluxes in the homogeneous model by as much as 10-20%. Differences of this magnitude can be therapeutically significant, emphasizing the need for proper validation of simplified treatment planning models.

It is well known that magnetic flux pileup can significantly speed up the rate of magnetic reconnection in high Lundquist number resistive MHD,allowing reconnection to proceed at a rate which is insensitive to the plasma resistivity over a wide range of Lundquist number. Hence, pileup is a possible solution to the Sweet-Parker time scale problem. Unfortunately, pileup tends to saturate above a critical value of the Lundquist number, S_c, where the value ofS_c depends on initial and boundary conditions, with Sweet-Parker scaling returning above S_c. It has been argued (see Dorelli and Bim [2003] and Dorelli [2003]) that the Hall effect can allow flux pileup to saturate (when the scale of the current sheet approaches ion inertial scale, di) before the reconnection rate begins to stall. However, the resulting saturated reconnection rate, while insensitive to the plasma resistivity, was found to depend strongly on the di. In this presentation, we revisit the problem of magnetic island coalescence (which is a well known example of flux pileup reconnection), addressing the dependence of the maximum coalescence rate on the ratio of di in the "large island" limit in which the following inequality is always satisfied: l_eta di lambda, where I_eta is the resistive diffusion length and lambda is the island wavelength.

In a graphite pile, EDF or Inca type reactor, it is necessary to know the value of the intermediate neutronflux at the output of the lateral reflector in order to determine more precisely the neutronflux at the level of ionisation chambers. A sub critical pile of graphite and natural uranium was built, allowing to reconstitute the geometry of the radiation sources and the disposition of inferior and lateral protections of these piles. This pile is supplied with thermal neutrons coming from the Nereide light water type reactor. Some measurements of intermediate neutronflux have been made in this pile in order to establish a formalism for neutronflux calculation in slowing down in a whole core-lateral reflector, from the distribution of the thermal neutronsflux in the core. The flux calculation is done by age theory in three dimensions, in two homogenous media, separated by an axially semi infinite and laterally finite plane. One of these media includes a distribution of source. The constants are modified in order to take into account the presence of empty channels in the stacking. These calculations are done by the Malaga code. The checking of the formalism has been made in a greater complex geometry of these reactors that introduces an uncertainty factor in the comparison of results. We can however tell that we estimate correctly the variation of the intermediate neutronsflux in the core as well as its descending in a holed lateral reflector. The ratio between the calculation and the experiment is inferior to 2 or 3. Most of the time to a factor 2. [French] Dans une pile a graphite, du type EdF ou Inca, il est necessaire de connaitre la valeur du flux de neutrons intermediaires a la sortie du reflecteur lateral, afin de determiner avec plus de precision le flux de neutrons au niveau des chambres d'ionisation. Il a ete construit un empilement sous-critique, graphite uranium naturel, qui permet de reconstituer la geometrie des sources de rayonnement et la

ENFMS(ex-core neutronflux monitoring system) is divided to source range, intermediate range and power ranger in accordance with its range and the output signal measurements of that are carried out with BF{sub 3} counter, fission chamber. There have been lots of study to adopt the wide-range measurement method which use only fission chamber through the whole reactor power. To do that is needs extending the power measurement range which is covered by fission chamber to lower power range. In lower power range the effect of noise in signal is greater relatively than that of high power range. The existing signal processing method to measurement plant power range in ENFMS in which the individual neutronflux pulse can be countered as the reactor power increased is MSV (mean square voltage) measurement. In this paper the extended method from MSV (2nd moment) mode to 3rd moment to improve the discrimination between neutron signal and background noise was studied. The simulation was shown that accuracy of power measurement in ENFMS using the method mention above would be improved. 2 tabs., 10 figs., 18 refs. (Author) .new.

The Malaysian Nuclear Agency (MNA) research reactor, commissioned in 1982, is a TRIGA Mark II swimming pool type reactor. When the core configuration changed in June 2009, it became essential to re-determine such neutronflux parameters as thermal to epithermal neutronflux ratio (f), epithermal neutronflux shape factor ({alpha}), thermal neutronflux ({phi}{sub th}) and epithermal neutronflux ({phi}{sub epi}) in the irradiation positions of MNA research reactor in order to guarantee accuracy in the application of k{sub 0}-neutron activation analysis (k{sub 0}-NAA).The f and {alpha} were determined using the bare bi-isotopic monitor and bare triple monitor methods, respectively; Au and Zr monitors were utilized in present study. The results for four irradiation positions are presented and discussed in the present work. The calculated values of f and {alpha} ranged from 33.49 to 47.33 and -0.07 to -0.14, respectively. The {phi}{sub th} and the {phi}{sub epi} were measured as 2.03 x 10{sup 12} (cm{sup -2} s{sup -1}) and 6.05 x 10{sup 10} (cm{sup -2} s{sup -1}) respectively. These results were compared to those of previous studies at this reactor as well as to those of reactors in other countries. The results indicate a good conformity with other findings.

The presence of crust on the inner walls of metallic ducts impairs transportation because crust completely or partially hinders the passage of fluid to the processing unit and causes damage to equipment connected to the production line. Its localization is crucial. With the development of the electronic imaging system installed at the Argonauta/Nuclear Engineering Institute (IEN)/National Nuclear Energy Commission (CNEN) reactor, it became possible to visualize crust in the interior of metallic piping of small diameter using real-time neutron radiography images obtained with a low neutronflux. The obtained images showed the resistance offered by crust on the passage of water inside the pipe. No discrepancy of the flow profile at the bottom of the pipe, before the crust region, was registered. However, after the passage of liquid through the pipe, images of the disturbances of the flow were clear and discrepancies in the flow profile were steep. This shows that this technique added the assembled apparatus was efficient for the visualization of the crust and of the two-phase flows.

We studied electric impedance of SiO{sub 2} nanomaterial at its initial state and after being exposed to continuous neutron irradiation for up to 20 hours. In doing so we employed a flux of neutrons of 2x10{sup 13} n⋅cm{sup −2}s{sup −1} while the frequency and temperature ranges amounted to 0,09 – 2.3 MHz and 100 – 400 K correspondingly. Analysis in terms of the Cole-Cole expression revealed that with increasing irradiation period the polarization and relaxation times decrease as a result of combination of nanoparticles. Moreover, it is demonstrated that the electric conductivity of samples, on the other hand, increases with the increasing irradiation period. At low temperatures formations of clusters at three distinct states with different energies were resolved.

Our recent paper on the "Search for NeutronFlux Generation in a Plasma Discharge Electrolytic Cell" [1] has as main goal the validation of the experiment in Ref.[2]. As a follow-up, Ref.[3] moves a set of objections on our procedure and presents argumentations on why the experiments should not yield the same results. We collect here additional material and calculations that contribute to understanding the observed discrepancies. Furthermore we prove that the absence of signals from Indium activation detectors reported also for the experiment of Ref.[2] is a clear indication that neutron production does not occur. [1] R.Faccini et al arXiv:1310.4749 [2] D.Cirillo et al, Key Engineering Materials 495, 104 (2012). [3] A.Widom et al. arXiv:1311.2447

Beams of epithermal neutrons are being used in the development of boron neutron capture therapy for cancer. This report describes a design study in which 235U fission plates and moderators are used to produce an epithermal neutron beam with higher intensity and better quality than the beam currently in use at the Brookhaven Medical Research Reactor (BMRR). Monte Carlo calculations are used to predict the neutron and gamma fluxes and absorbed doses produced by the proposed design. Neutronflux measurements at the present epithermal treatment facility (ETF) were made to verify and compare with the computed results where feasible. The calculations indicate that an epithermal neutron beam produced by a fission-plate converter could have an epithermal neutron intensity of 1.2 x 10(10) n/cm2.s and a fast neutron dose per epithermal neutron of 2.8 x 10(-11) cGy.cm2/nepi plus being forward directed. This beam would be built into the beam shutter of the ETF at the BMRR. The feasibility of remodeling the facility is discussed.

Whenever tasks involving the use of a nuclear reactor are carried out, e.g., radioisotope production, activation analysis, neutrongraphy, etc., it is necessary to know the magnitude of the associate neutronflux. The Argonauta reactor operating in Rio de Janeiro, at Instituto de Engenharia Nuclear - IEN/CNEN, was submitted to some modifications in its core, which made necessary to measure again its new neutronic characteristics, not only in the core itself, but also at the irradiation pads. In this type of research reactor, the neutrons are energetically distributed from values below 1 eV, to values reaching the magnitude of MeV. Therefore, depending on the kind of experiment to be conducted, it may become necessary to know the integrated neutronflux within certain energy ranges. In this work, the neutronflux for thermal and epithermal regions were determined by using the foil activation method. To accomplish this goal, two different techniques were applied. In the first technique {beta}-{gamma} gamma coincidence measurements were performed using a proportional 4{pi}{beta} gaseous detector and a NaI(Tl) scintillation detector, while in the second one, gamma spectroscopy was carried out using Hp-Ge and NaI(Tl) detectors. In both cases, the flux was computed using the FLUXO software, specially developed for this purpose. (author)

The demand of new and high precision cross section data for neutron-induced reactions is continuously growing, driven by the requirements from several fields of fundamental physics, as well as from nuclear technology, medicine, etc. Several neutron facilities are operational worldwide, and new ones are being built. In the coming years, neutron beam intensities never reached up to now will be available, thus opening new scientific and technological frontiers. Among existing facilities, n_TOF at CERN provides a high intensity pulsed neutron beam in a wide energy range (thermal to GeV) and with an extremely competitive energy resolution that also allows spectroscopy studies. In order to ensure high quality measurements, the neutron beams must be fully characterized as a function of the neutron energy, in particular by measuring the neutronflux and the beam transverse profile with high accuracy. In 2014 a new experimental area (EAR2), with a much higher neutronflux, has been completed and commissioned at n_TOF. In order to characterize the neutron beam in the newly built experimental area at n_TOF, two suitable diagnostics devices have been built by the INFN-LNS group. Both are based on silicon detectors coupled with 6Li converter foils, in particular Single Pad for the flux measurement and Position Sensitive (strips and others) for the beam profile. The devices have been completely characterized with radioactive sources and with the n_TOF neutron beam, fulfilling all the specifications and hence becoming immediately operational. The performances of these devices and their high versatility, in terms of neutron beam intensity, make them suitable to be used in both n_TOF experimental areas. A description of the devices and the main results obtained so far will be presented.

The activation of the biological shield is an important process for waste management considerations of nuclear facilities. The final activity can be estimated by modeling using the neutronflux density rather than the radiometric approach of activity measurements. Measurement series at the TRIGA reactor Vienna reveal that the flux density next to the biological shield is in the order of 10{sup 9} cm{sup -2} s{sup -1} at maximum power; but it is strongly influenced by reactor installations. The data allow the estimation of the final waste categorization of the concrete according to the Austrian legislation. - Highlights: > Neutron activation is an important process for the waste management of nuclear facilities. > Biological shield of the TRIGA reactor Vienna has been topic of investigation. > Flux values allow a categorization of the concrete concerning radiation protection legislation. > Reactor installations are of great importance as neutron sources into the biological shield. > Every installation shows distinguishable flux profiles.

Ion cyclotron resonance heating experiments using antenna, in the high Reid side (HFS) have been carried out on HT-7 in different target plasmas. Unlike a standard-mode conversion heating scheme with dominant electron heating, anomalous ion heating and DD neutronfluxes higher than those estimated from thermal ions were observed in the present experiments with the ion-ion hybrid resonant layer near the center of plasma. The features of ion cyclotron range frequency (ICRF) antenna in HFS and experiments suggest that this is most probably due to the nonlinear 3/2 harmonic deuterium heating by the mode-converted ion Bernstein wave, which could produce a high energy tail on ion energy distribution.%Ion cyclotron resonance heating experiments using antenna in the high field side (HFS) have been carried out on HT-7 in different target plasmas.Unlike a standard-mode conversion heating scheme with dominant electron heating,anomalous ion heating and DD neutronfluxes higher than those estimated from thermal ions were observed in the present experiments with the ion-ion hybrid resonant layer near the center of plasma.The features of ion cyclotron range frequency (ICRF) antenna in HFS and experiments suggest that this is most probably due to the nonlinear 3/2 harmonic deuterium heating by the mode-converted ion Bernstein wave,which could produce a high energy tail on ion energy distribution.Neutron diagnostics have been applied in ion cyclotron range frequency (ICRF) plasmas on HT-7 for measurements of the fusion reaction product,which give a direct measure of the ICRF heating.The neutron emission is recorded by a 3He proportional counter,whose sensitive size is φ30 mm × 300 mm,gas pressure is 49.34 kPa and the responsibility to thermal neutrons is 133 cps/n.cm-2.s-1.It exploits large reaction cross sections and is therefore embedded in polythene moderators to thermalize the incident neutrons.

It has been reviewed the frequency dependencies of electrical conductivity of nanoparticles affected by neutronflux at different times and initial state, at various constant temperatures such as 100, 200, 300 and 400 K. Measurements have been carried out at each temperature at the different 97 values of frequency in the 1 Hz - 1 MHz range. From interdependence between real and imaginary parts of electrical conductivity it has been determined the type of conductivity. Moreover, in the work it is given the mechanism of electrical conductivity according to the obtained results. (Author)

A neutronflux monitor .(NFM) is a key diagnostic system in the International Ther- monuclear Experimental Reactor (ITER), and may provide readings of a series of important parameters in fusion reaction processes. As a valuable part of the main electronics system of the NFM, the high-speed baseline restorer we designed is an important signal conversion plug-in which can restore the input signal baseline offset to a zero level, while keeping the output pulse signal waveform from the preamplifier basically unchanged.

The depolarization of a neutron beam passing through a sample of the high-{ital T}{sub {ital c}} superconductor YBa{sub 2}Cu{sub 3}O{sub 7{minus}{delta}} has been measured as a function of temperature and applied field. The difference in behavior between field-cooled and zero-field-cooled states, the observation of hysteresis correlated with {ital H}{sub {ital c}1}, and the disappearance of the effect near 55 K (below {ital T}{sub {ital c}}) suggest an explanation in terms of vortex line lattice formation with possible connection to recently proposed flux line entanglement and melting.

Small-angle neutron diffraction measurements from the flux lattice in a single crystal of the cubic Laves` phase superconductor, CeRu{sub 2}, are reported. The mixed state is described in terms of aligned rigid bundles of vortices. The bundle diameters decrease above 1/2H{sub c2} (consistent with collective weak pinning theory) and become comparable with the penetration length at a field at which a `peak effect` is seen in magnetisation measurements. A clear memory of field histories that pass through the `peak effect` region is also found; however, some of the induced disorder can be removed by subsequently cycling the field. (orig.).

Aeroball system is attractive in several aspects because it can easily transport the map of neutronflux distribution to be measured from incore to outside of a reactor vessel.However,before the aeroball system is put to practical use in the heating reactor.there are four topics that have to be further studied.They are the stability of the activated positions,enhancement of signal/noise(S/N)ratio,distributed control and data-acquisition system and on-lin nbeutron flux distribution reconstruction.Besides describing the rasons for them,this paper gives out the theory,concept and solution about the first two topics and it is helptul to give the possibility to enhance the reactor-power.

In this work, an analytical discrete ordinates method is used to solve a nodal formulation of a neutron transport problem in x, y-geometry. The proposed approach leads to an important reduction in the order of the associated eigenvalue systems, when combined with the classical level symmetric quadrature scheme. Auxiliary equations are proposed, as usually required for nodal methods, to express the unknown fluxes at the boundary introduced as additional unknowns in the integrated equations. Numerical results, for the problem defined by a two-dimensional region with a spatially constant and isotropically emitting source, are presented and compared with those available in the literature. (authors)

Polarized neutron radiography was used to study the three-dimensional magnetic flux distribution inside of single-crystal and polycrystalline Pb cylinders with large (cm3) volume and virtually zero demagnetization. Experiments with single crystals being in the Meissner phase (Tflux was nearly constant. The expelled field outside of the samples followed 1/R dependence. These measurements provided a unique and detailed picture of macroscopic superconducting samples, confirming the existence of both uniform bulk Meissner expulsion in single crystals and bulk flux trapping with nearly-Bean-model profiles due to flux pinning in polycrystalline samples.

The capability to measure bursts of neutron fluences and gamma/x-ray fluxes directly with charge coupled device (CCD) cameras while being able to distinguish between the video signals produced by these two types of radiation, even when they occur simultaneously, has been demonstrated. Volume and area measurements of transient radiation-induced pixel charge in English Electric Valve (EEV) Frame Transfer (FT) charge coupled devices (CCD's) from irradiation with pulsed neutrons (14 MeV) and Bremsstrahlung photons (4-12 MeV endpoint) are utilized to calibrate the devices as radiometric imaging sensors capable of distinguishing between the two types of ionizing radiation. Measurements indicate approx. = .05 V/rad responsivity with greater than or = 1 rad required for saturation from photon irradiation. Neutron-generated localized charge centers or 'peaks' binned by area and amplitude as functions of fluence in the 105 to 107 n/cc range indicate smearing over approx. 1 to 10 percent of the CCD array with charge per pixel ranging between noise and saturation levels.

The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) is currently undergoing an upgrading program, a part of which is to increase the diameters of two of the four radiation beam tubes (HB-2 and HB-4). This change will cause increased neutron and gamma radiation dose rates at and near locations where the tubes penetrate the vessel wall. Consequently, the rate of radiation damage to the reactor vessel wall at those locations will also increase. This report summarizes calculations of the neutron and gamma flux (particles/cm{sup 2}/s) and the dpa rate (displacements/atom/s) in iron at critical locations in the vessel wall. The calculated dpa rate values have been recently incorporated into statistical damage evaluation codes used in the assessment of radiation induced embrittlement. Calculations were performed using models based on the discrete ordinates methodology and utilizing ORNL two-dimensional and three-dimensional discrete ordinates codes. Models for present and proposed beam tube designs are shown and their results are compared. Results show that for HB-2, the dpa rate in the vessel wall where the tube penetrates the vessel will be increased by {approximately}10 by the proposed enlargement. For HB-4, a smaller increase of {approximately}2.6 is calculated.

Full Text Available This work was performed within the international project “Energy plus Transmutation of Radioactive Wastes” (E&T - RAW for investigations of energy production and transmutation of radioactive waste of the nuclear power industry. 89Y (Yttrium 89 samples were located in the Quinta assembly in order to measure an average high neutronflux density in three different energy ranges using deuteron and proton beams from Dubna accelerators. Our analysis showed that the neutron density flux for the neutron energy range 20.8 - 32.7 MeV is higher than for the neutron energy range 11.5 - 20.8 MeV both for protons with an energy of 0.66 GeV and deuterons with an energy of 2 GeV, while for deuteron beams of 4 and 6 GeV we did not observe this.

Five different central difference schemes, based on a conservative differencing form of the Kennedy and Gruber skew-symmetric scheme, were compared with six different upwind schemes based on primitive variable reconstruction and the Roe flux. These eleven schemes were tested on a one-dimensional acoustic standing wave problem, the Taylor-Green vortex problem and a turbulent channel flow problem. The central schemes were generally very accurate and stable, provided the grid stretching rate was kept below 10%. As near-DNS grid resolutions, the results were comparable to reference DNS calculations. At coarser grid resolutions, the need for an LES SGS model became apparent. There was a noticeable improvement moving from CD-2 to CD-4, and higher-order schemes appear to yield clear benefits on coarser grids. The UB-7 and CU-5 upwind schemes also performed very well at near-DNS grid resolutions. The UB-5 upwind scheme does not do as well, but does appear to be suitable for well-resolved DNS. The UF-2 and UB-3 upwind schemes, which have significant dissipation over a wide spectral range, appear to be poorly suited for DNS or LES.

We have tested the Monte Carlo code FLUKA for its ability to assist in the development of a better system for the in vivo measurement of fluorine. We used it to create a neutronflux map of the inside of the in vivo neutron activation analysis irradiation cavity at the McMaster Accelerator Laboratory. The cavity is used in a system that has been developed for assessment of fluorine levels in the human hand. This study was undertaken to (i) assess the FLUKA code, (ii) find the optimal hand position inside the cavity and assess the effects on precision of a hand being in a non-optimal position and (iii) to determine the best location for our γ-ray detection system within the accelerator beam hall. Simulation estimates were performed using FLUKA. Experimental measurements of the neutronflux were performed using Mn wires. The activation of the wires was measured inside (1) an empty bottle, (2) a bottle containing water, (3) a bottle covered with cadmium and (4) a dry powder-based fluorine phantom. FLUKA was used to simulate the irradiation cavity, and used to estimate the neutronflux in different positions both inside, and external to, the cavity. The experimental results were found to be consistent with the Monte Carlo simulated neutronflux. Both experiment and simulation showed that there is an optimal position in the cavity, but that the effect on the thermal flux of a hand being in a non-optimal position is less than 20%, which will result in a less than 10% effect on the measurement precision. FLUKA appears to be a code that can be useful for modeling of this type of experimental system.

A large volume (1m3) spherical proportional counter has been developed at CEA/Saclay, for low fluxneutron measurements. The high voltage is applied to a small sphere 15mm in diameter, located in the center of the counter and the wall of the counter is grounded. Neutrons can be measured successfully, with high sensitivity, using 3He gas in the detector. The proton and tritium energy deposition in the drift gaseous volume, from the reaction 3He(n,p)3H, can provide the neutron spectra from thermal neutrons up to several MeV. The detector has been installed in the underground laboratory in Modane (LSM) to measure the neutron background. The sphere has been has been filled with gas mixture of Ar + 2% CH4 +3gr He-3, at 275 mbar. The thermal neutron peak is well separated from the cosmic ray and gamma background, permitting of neutronflux calculation. Other potential applications requiring large volume of about 10 m in radius are described in detail in reference

In this paper, we present an accurate and robust scaling operator based on material optical thickness (OT) for the least-squares spherical harmonics (LSP$_N$) method for solving neutron transport problems. LSP$_N$ without proper scaling is known to be erroneous in highly scattering medium, if the optical thickness of the material is large. A previously presented scaling developed by Manteuffel, et al.\\ does improve the accuracy of LSP$_N$, in problems where the material is optically thick. With the method, however, essentially no scaling is applied in optically thin materials, which can lead to an erroneous solution with presence of highly scattering medium. Another scaling approach, called the reciprocal-removal (RR) scaled LSP$_N$, which is equivalent to the self-adjoint angular flux (SAAF) equation, has numerical issues in highly-scattering materials due to a singular weighting. We propose a scaling based on optical thickness that improves the solution in optically thick media while avoiding the singularit...

This report summarized several irradiations that have been made to determine the neutronflux distributions in the core of the JEN-1 reactor. Gold foils of 380 {mu} gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs.

The neutronflux monitoring system of the French GEN-IV sodium-cooled fast reactor will rely on high temperature fission chambers installed in the reactor vessel and capable of operating over a wide-range neutronflux. The definition of such a system is presented and the technological solutions are justified with the use of simulation and experimental results. (authors)

1x1 cm sup 2 silicon pad p sup + -n-n sup + detectors were irradiated with fast neutrons from the TRIGA research reactor in Ljubljana to fluences from 5x10 sup 1 sup 3 to 10 sup 1 sup 4 n/cm sup 2. The observed time development of annealing of the full-depletion voltage (FDV) could be fitted by a constant and two exponentials. The characteristic time of the fast component is 4 h, independent of temperature in the interval 0-15 deg. C. A comparison of MESA and planar pad detectors shows a 20-30% lower FDV for the MESA. A search for a flux dependence of the radiation damage was performed in the range from 2x10 sup 8 to 5x10 sup 1 sup 5 n/cm sup 2 s and no systematic differences were observed.

With base in the study of the real signals of neutronflux of instability events occurred in the Laguna Verde nuclear power plant where the nucleus oscillation phenomena of the reactor are in the 0 to 2.5 Hz range, it has been seen the possibility about the development a surveillance and diagnostic equipment capable to analyze in real time the behavior of nucleus in this frequencies range. An important method for surveillance the stability of the reactor nucleus is the use of the Power spectral density which allows to determine the frequencies and amplitudes contained in the signals. It is used an instrument carried out by LabVIEW graphic programming with a data acquisition card of 16 channels which works at Windows 95/98 environment. (Author)

The Mini-INCA project is dedicated to the measurement of incineration-transmutation chains and potentials of minor actinides in high-intensity thermal neutronfluxes. In this context, new types of detectors and methods of analysis have been developed. The {sup 241}Am and {sup 232}Th transmutation-incineration chains have been studied and several capture and fission cross sections measured very precisely, showing some discrepancies with existing data or evaluated data. An impact study was made on different based-like GEN-IV reactors. It underlines the necessity to proceed to precise measurements for a large number of minor-actinides that contribute to these future incineration scenarios. (authors)

A facility for the irradiation of a section of patients' explanted liver and lung was constructed at RA-3 reactor, Comision Nacional de Energia Atomica, Argentina. The facility, located in the thermal column, is characterized by the possibility to insert and extract samples without the need to shutdown the reactor. In order to reach the best levels of security and efficacy of the treatment, it is necessary to perform an accurate dosimetry. The possibility to simulate neutronflux and absorbed dose in the explanted organs, together with the experimental dosimetry, allows setting more precise and effective treatment plans. To this end, a computational model of the entire reactor was set-up, and the simulations were validated with the experimental measurements performed in the facility.

We have measured neutronflux distribution around the core tank of STACY heterogeneous core by position sensitive proportional counter (PSPC) to develop the method to measure reactivity for subcritical systems. The neutronflux distribution data in the position accuracy of +-13 mm have been obtained in the range of uranium concentration of 50g/L to 210g/L both in critical and in subcritical state. The prompt neutron decay constant, alpha, was evaluated from the measurement data of pulsed neutron source experiments. We also calculated distribution of neutronflux and sup 3 He reaction rates at the location of PSPC by using continuous energy Monte Carlo code MCNP. The measurement data was compared with the calculation results. As results of comparison, calculated values agreed generally with measurement data of PSPC with Cd cover in the region above half of solution height, but the difference between calculated value and measurement data was large in the region below half of solution height. On the other hand, ...

We have measured neutronflux distribution around the core tank of STACY heterogeneous core by position sensitive proportional counter (PSPC) to develop the method to measure reactivity for subcritical systems. The neutronflux distribution data in the position accuracy of {+-}13 mm have been obtained in the range of uranium concentration of 50g/L to 210g/L both in critical and in subcritical state. The prompt neutron decay constant, {alpha}, was evaluated from the measurement data of pulsed neutron source experiments. We also calculated distribution of neutronflux and {sup 3}He reaction rates at the location of PSPC by using continuous energy Monte Carlo code MCNP. The measurement data was compared with the calculation results. As results of comparison, calculated values agreed generally with measurement data of PSPC with Cd cover in the region above half of solution height, but the difference between calculated value and measurement data was large in the region below half of solution height. On the other hand, calculated value agreed well with measurement data of PSPC without Cd cover. (author)

The EURISOL (The EURopean Isotope Separation On-Line Radioactive Ion Beam) project aims atproducing high intensity radioactive ion beams produced by neutron-induced fission on fissile targets(235U) surrounding a liquid mercury converter. A proton beam of 1GeV and 4MW impinges on theconverter, generating, by spallation reactions, high neutronfluxes that induce fission in thesurrounding fissile targets.In this work the state-of-the-art Monte Carlo codes MCNPX and FLUKA were used to assess theneutronics performance of the system, which geometry, inspired in the MAFF concept, allows aversatile manipulation of the fission targets. The first objective of the study was to optimize thegeometry and the materials used in the fuel and reflector elements of the system, in order to achievethe highest possible fission rates. Indeed, it is shown that the appropriate combination of fission targetmaterial and surrounding reflector material leads to the aimed value of 1015 fissions/s per fissiontarget. The second part of this...

The Th-U cycle was studied through the evolution of a 100 μg 232Th sample irradiated in a moderated neutronflux of 8.010 14 n/cm 2/s, intensity close to that of a thermal molten salt reactor. After 43 days of irradiation and 6 months of cooling, a precise mass spectrometric analysis, using both TIMS and MC-ICP-MS techniques, was performed, according to a rigorous methodology. The measured thorium and uranium isotopic ratios in the final irradiated sample were then compared with integral simulations based on evaluated data; an overall good agreement was seen. Four important thermal neutron-capture cross-sections were also extracted from the measurements, 232Th (7.34±0.21 b), 233Pa (38.34±1.78 b), 234U (106.12±3.34 b), and 235U (98.15±11.24 b). Our 232Th and 235U results confirmed existing values whereas the cross-sections of 233Pa and 234U (both key parameters) have been redefined.

Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

The Lunar Reconnaissance Orbiter (LRO) was launched June 18, 2009 during an historic space-age era of minimum solar activity [1]. The lack of solar sunspot activity signaled a complex set of heliospheric phenomena [2,3,4] that also gave rise to a period of unprecedentedly high Galactic Cosmic Ray (GCR) flux [5]. These events coincided with the primary mission of the Lunar Exploration Neutron Detector (LEND, [6]), onboard LRO in a nominal 50-km circular orbit of the Moon [7]. Methods to calculate the emergent neutron albedo population using Monte Carlo techniques [8] rely on an estimate of the GCR flux and spectra calibrated at differing periods of solar activity [9,10,11]. Estimating the actual GCR flux at the Moon during the LEND's initial period of operation requires a correction using a model-dependent heliospheric transport modulation parameter [12] to adjust the GCR flux appropriate to this unique solar cycle. These corrections have inherent uncertainties depending on model details [13]. Precisely determining the absolute neutron and GCR fluxes is especially important in understanding the emergent lunar neutrons measured by LEND and subsequently in estimating the hydrogen/water content in the lunar regolith [6]. LEND is constructed with a set of neutron detectors to meet differing purposes [6]. Specifically there are two sets of detector systems that measure the flux of epithermal neutrons: a) the uncollimated Sensor for Epi-Thermal Neutrons (SETN) and b) the Collimated Sensor for Epi-Thermal Neutrons (CSETN). LEND SETN and CSETN observations form a complementary set of simultaneous measurements that determine the absolute scale of emergent lunar neutronflux in an unambiguous fashion and without the need for correcting to differing solar-cycle conditions. LEND measurements are combined with a detailed understanding of the sources of instrumental back-ground, and the performance of CSETN and SETN. This comparison allows us to calculate a constant scale factor

A new neutron-cooling mechanism is proposed with potential benefits for novel intense sources of very cold neutrons with wavelengths >2 nm, and for enhancing the production of ultracold neutrons. It employs inelastic magnetic scattering in weakly absorbing, cold paramagnetic systems. Kinetic energy is removed from the neutron stepwise in constant decrements determined by the Zeeman energy of paramagnetic atoms or ions in an external magnetic field, or by zero-field level splittings in magnetic molecules. The stationary neutron transport equation is analyzed for an infinite, homogeneous medium with Maxwellian neutron sources, using inelastic scattering cross sections derived in an appendix. Nonmagnetic inelastic scattering processes are neglected. The solution therefore still underestimates very cold neutron densities that should be achievable in a real medium. Molecular oxygen with its triplet ground state appears particularly promising, notably as a host in fully deuterated O2-clathrate hydrate. Other possibilities are dry O2-4He van der Waals clusters and O2 intercalated in fcc-C60. For conversion of cold to ultracold neutrons, where an incident neutron imparts only a single energy quantum to the medium, the paramagnetic scattering in the clathrate system is found to be stronger, by more than an order of magnitude, than the single-phonon emission in superfluid helium, when evaluated for an incident neutron spectrum with the optimum temperature for the respective medium. Moreover, the multistep paramagnetic cooling cascade leads to further strong enhancements of very cold neutron densities, e.g., by a factor 14 (57) for an initial neutron temperature of 30 K (100 K ), for the moderator held at about 1.3 K . Due to a favorable Bragg cutoff of the O2 clathrate, the cascade-cooling can take effect in a moderator with linear extensions smaller than a meter.

This treatise develops new methods for high fluxneutron radiography and high fluxneutron tomography and describes some of their applications in actual experiments. Instead of single images, time series can be acquired with short exposure times due to the available high intensity. To best use the increased amount of information, new estimators are proposed, which extract accurate results from the recorded ensembles, even if the individual piece of data is very noisy and in addition severely affected by systematic errors such as an influence of gamma background radiation. The spatial resolution of neutron radiographies, usually limited by beam divergence and inherent resolution of the scintillator, can be significantly increased by scanning the sample with a pinhole-micro-collimator. This technique circumvents any limitations in present detector design and, due to the available high intensity, could be successfully tested. Imaging with scattered neutrons as opposed to conventional total attenuation based imaging determines separately the absorption and scattering cross sections within the sample. For the first time even coherent angle dependent scattering could be visualized space-resolved. New applications of high fluxneutron imaging are presented, such as materials engineering experiments on innovative metal joints, time-resolved tomography on multilayer stacks of fuel cells under operation, and others. A new implementation of an algorithm for the algebraic reconstruction of tomography data executes even in case of missing information, such as limited angle tomography, and returns quantitative reconstructions. The setup of the world-leading high flux radiography and tomography facility at the Institut Laue-Langevin is presented. A comprehensive appendix covers the physical and technical foundations of neutron imaging. (orig.)

In this dissertation, a fast neutronflux-monitoring channel, which is based on the use of SiC semiconductor detectors is designed, modeled and experimentally evaluated as a power monitor for the Gas Turbine Modular Helium Reactors. A detailed mathematical model of the SiC diode detector and the electronic processing channel is developed using TRIM, MATLAB and PSpice simulation codes. The flux monitoring channel is tested at the OSU Research Reactor. The response of the SiC neutron-monitoring channel to neutrons is in close agreement to simulation results. Linearity of the channel response to thermal and fast neutronfluxes, pulse height spectrum of the channel, energy calibration of the channel and the detector degradation in a fast neutronflux are presented. Along with the model of the neutron monitoring channel, a Simulink model of the GT-MHR core has been developed to evaluate the power monitoring requirements for the GT-MHR that are most demanding for the SiC diode power monitoring system. The Simulink model is validated against a RELAP5 model of the GT-MHR. This dyanamic model is used to simulate reactor transients at the full power and at the start up, in order to identify the response time requirements of the GT-MHR. Based on the response time requirements that have been identified by the Simulink model and properties of the monitoring channel, several locations in the central reflector and the reactor cavity are identified to place the detector. The detector lifetime and dynamic range of the monitoring channel at the detector locations are calculated. The channel dynamic range in the GT-MHR central reflector covers four decades of the reactor power. However, the detector does not survive for a reactor refueling cycle in the central reflector. In the reactor cavity, the detector operates sufficiently long; however, the dynamic range of the channel is smaller than the dynamic range of the channel in the central reflector.

The core restraints of advanced gas-cooled reactors are important structural components that are required to maintain the geometric integrity of the cores. A review of neutron dosimetry for the sister stations Hunterston B and Hinkley Point B identified that earlier conservative assessments predicted high thermal neutron dose rates to key components of the restraint structure (the restraint rod welds), with the implication that some of them may be predicted to fail during a seismic event. A revised assessment was therefore undertaken [Thornton, D. A., Allen, D. A., Tyrrell, R. J., Meese, T. C., Huggon, A.P., Whiley, G. S., and Mossop, J. R., 'A Dosimetry Assessment for the Core Restraint of an Advanced Gas Cooled Reactor,' Proceedings of the 13. International Symposium on Reactor Dosimetry (ISRD-13, May 2008), World Scientific, River Edge, NJ, 2009, W. Voorbraak, L. Debarberis, and P. D'hondt, Eds., pp. 679-687] using a detailed 3D model and a Monte Carlo radiation transport program, MCBEND. This reassessment resulted in more realistic fast and thermal neutron dose recommendations, the latter in particular being much lower than had been thought previously. It is now desirable to improve confidence in these predictions by providing direct validation of the MCBEND model through the use of neutronflux measurements. This paper describes the programme of work being undertaken to deploy two neutronflux measurement 'stringers' within the side-core region of one of the Hunterston B reactors for the purpose of validating the MCBEND model. The design of the stringers and the determination of the preferred deployment locations have been informed by the use of detailed MCBEND flux calculations. These computational studies represent a rare opportunity to design a flux measurement beforehand, with the clear intention of minimising the anticipated uncertainties and obtaining measurements that are known to be representative of the neutron fields to which

A sub-critical advanced reactor based on Tokamak technology with a D-T fusion neutron source is an innovative type of nuclear system. Due to the large number of neutrons produced by fusion reactions, such a system could be useful in the transmutation process of transuranic elements (Pu and minor actinides (MAs)). However, to enhance the MA transmutation efficiency, it is necessary to have a large neutron wall loading (high neutron fluence) with a broad energy spectrum in the fast neutron energy region. Therefore, it is necessary to know and define the neutron fluence along the radial axis and its characteristics. In this work, the neutronflux and the interaction frequency along the radial axis are evaluated for various materials used to build the first wall. W alloy, beryllium, and the combination of both were studied, and the regions more suitable to transmutation were determined. The results demonstrated that the best zone in which to place a transmutation blanket is limited by the heat sink and the shield block. Material arrangements of W alloy/W alloy and W alloy/beryllium would be able to meet the requirements of the high fluence and hard spectrum that are needed for transuranic transmutation. The system was simulated using the MCNP code, data from the ITER Final Design Report, 2001, and the Fusion Evaluated Nuclear Data Library/MC-2.1 nuclear data library.

In the framework of nuclear waste transmutation studies, the Mini-INCA project has been initiated at CEA/DSM with objectives to determine optimal conditions for transmutation and incineration of minor actinides (MA) in high intensity neutronfluxes. Our experimental tools based on alpha- and gamma-spectroscopy of the samples and the development of micro fission chambers could gather either microscopic information on nuclear reactions (total or partial cross sections for neutron capture and/or fission reactions) or macroscopic information on transmutation and incineration potentials. Neutron capture cross sections of selected actinides ({sup 241}Am, {sup 242}Am, {sup 242}Pu, {sup 237}Np) have already been measured at ILL, showing some discrepancies when compared to evaluated data libraries but in overall good agreement with recent data. The studies and possibilities offer by the MEGAPIE project to assess neutronic performances of a 1 MW spallation target and the incineration of MA in a representative neutronflux of a spallation source are also discussed. (orig.)

The production of radioisotopes is one of the most important applications of nuclear research reactors. This study investigated a method called Flux Trap, which is used to increase the yield of production of radioisotopes in nuclear reactors. The method consists in the rearrangement of the fuel rods to allow the increase of the thermal neutronflux in the irradiation region inside the reactor core, without changing the standard reactor power level. Various configurations were assembled with the objective of finding the configuration with the highest thermal neutronflux in the region of irradiation. The method of activation analysis was used to measure the thermal neutronflux and determine the most efficient reactor core configuration . It was found that there was an increase in the thermal neutronflux of 337% in the most efficient configuration, which demonstrates the effectiveness of the method. (author)

New and exciting results for measuring neutron star surface temperatures began with the successful launch of the Chandra X-ray observatory. Among these results are new detections of neutron star surface temperatures which have made it possible to seriously test neutron star thermal evolution theories. The important new temperature determination of the Vela pulsar (Pavlov, et al., 2001a) requires a non-standard cooling scenario to explain it. Apart from this result, we have measured PSR B1055-52's surface temperature in this thesis, determining that it can be explained by standard cooling with heating. Our spectral fit of the combined data from ROSAT and Chandra have shown that a three component model, two thermal blackbodies and an non-thermal power-law, is required to explain the data. Furthermore, our phase resolved spectroscopy has begun to shed light on the geometry of the hot spot on PSR B1055-52's surface as well as the structure of the magnetospheric radiation. Also, there is strong evidence for a thermal distribution over its surface. Most importantly, the fact that PSR B1055-52 does not have a hydrogen atmosphere has been firmly established. To reconcile these two key observations, on the Vela pulsar and PSR B1055-52, we tested neutron star cooling with neutrino processes including the Cooper pair neutrino emission process. Overall, it has been found that a phase change associated with pions being present in the cores of more massive neutron stars explains all current of the data. A transition from neutron matter to pion condensates in the central stellar core explains the difference between standard and non-standard cooling scenarios, because the superfluid suppression of pion cooling will reduce the emissivity of the pion direct URCA process substantially. A neutron star with a mass of [Special characters omitted.] with a medium stiffness equation of state and a T72 type neutron superfluid models the standard cooling case well. A neutron star of [Special

A geologic analysis of 274 images acquired by the high-resolution MOC camera onboard the Mars Global Surveyor spacecraft within the Arabia Terra low neutronflux anomaly (which is indicative of an anomalously high abundance of hydrogen: up to 16 wt % of the equivalent amount of water) was performed. Correlation between the enhanced abundance of equivalent water with the presence of dust on the surface was found. Since dust plays a key role in condensation of water from the atmosphere, we suppose that the anomalies could result from the retention of atmospheric moisture. To analyze this suggestion, we performed a theoretical modeling that allowed us to map the planetary-scale distributions of several meteorological parameters responsible for the atmospheric moisture condensation. Two antipodal regions coinciding rather well with the Arabia Terra anomaly and the geographically antipodal anomaly southwest of Olympus Mons were found in the maps. This suggests that the anomalies are rather recent than ancient formations. They were probably formed by a sink of moisture from the atmosphere in the areas where present meteorological conditions support this sink. Geological parameters, primarily the presence of dust, only promote this process. We cannot exclude the possibility that the Martian cryosphere, rather than the atmosphere, supplied the studied anomalies with moisture during their formation: the thermodynamic conditions in the anomaly areas could block the moisture flux from the Martian interior in the upper regolith layer. The moisture coming from the atmosphere or from the interior is likely held as chemically bound water entering into the structure of water-bearing minerals (probably, hydrated magnesium sulfates) directly from the vapor; or the moisture precipitates as frost, penetrates into microfissures, and then is bound in minerals. Probably, another geologic factor—the magnesium sulfate abundance—works in the Arabia Terra anomaly.

The spatial dependence of thermal and fast neutronflux was measured axially in the core of a 1 MW research reactor. The measurements were made by a thin optical fibre detector with a neutron sensitive ZnS(Ag) scintillation tip. For thermal neutrons sup 6 Li was used, whereas for fast neutrons sup 2 sup 3 sup 2 Th was used as neutron converter. The spatial dependence was measured by moving the fibre axially with a uniform speed. The measurement takes a few minutes, compared to up to 10 h with the conventional wire activation method. Comparison with traditional measurements shows a good agreement. (author)

This study aims to determine the distribution of thermal neutronflux in the IPEN/MB-01 nuclear reactor core assembled with cylindrical core configuration of minor excess of reactivity with 568 fuel rods (28 fuel rods in diameter). The thermal neutronflux at the positions of irradiation derive from the method of reaction rate using gold foils. The experiment consists in inserting gold activations foils with and without cadmium coverage (cadmium boxes with 0.0502 cm thickness) in several positions throughout the active core. After irradiation, activity induced by nuclear reaction rates over gold foils is assessed by gamma ray spectrometry using a high-purity germanium (HPGe) detector. Experimental results are compared to those derived from calculations performed using a three dimensional CITATION diffusion code and MCNP-5 code and a proper nuclear data library. While calculated neutronflux data shows good agreement with experimental values in regions with little disturbance in the neutronflux, also showing that in the region of the reflectors of neutrons and near the control rods, the diffusion theory is not very precise. The average value of thermal neutronflux obtained experimentally compared to the calculated value by CITATION code and MCNP-5 code respectively show a difference of 1.18% and 0.84% at a nuclear power level of 74.65 ± 3.28 % watts. The average measured value of thermal neutronflux is 4.10 10{sup 8} ± 5.25% n/cm{sup 2}s. (author)

We analyze the latest Suzaku observation of the bright neutron star system Serpens X-1 taken in 2013 October and 2014 April. The observation was taken using the burst mode and only suffered mild pile-up effects. A broad iron line is detected in the X-ray spectrum. We test different models and find that the iron line is asymmetric and best interpreted by relativistic reflection. The relativistically broadened iron line is generally believed to originate from the innermost regions of the accretion disk, where strong gravity causes a series of special and general relativistic effects. The iron line profile indicates an inner radius of ~8 $R_{\\rm G}$. The asymmetric iron line has been observed in a number of previous observations, which gives several inner radius measurements at different flux states. We find the inner radius to be consistent over a wide range of luminosity, implying that the inner radius of Serpens X-1 does not evolve significantly over the range of $L/L_{\\rm Edd}$ ~ 0.2-0.6.

Instrumental Neutron Activation Analysis using k0 (k0-INAA) method has been used to determine a number of elements in sediment samples collected from El-Manzala Lake in Egypt. k0-INAA according to Westcott's formalism has been implemented using the complete irradiation kit of the fast pneumatic rabbit and some selected manually loaded irradiation sites for short and long irradiation at Egypt Second Research Reactor (ETRR-2). Zr-Au and Co sets as neutronflux monitors are used to determine the neutronflux parameters (f and α) in each irradiation sites. Two reference materials IAEA Soil-7 samples have been inserted and implemented for data validation and an internal monostandard multi monitor used (k0 based IM-NAA). It was given a good agreement between the experimental analyzed values and that obtained of the certified values. The major and trace elements in the sediment samples have been evaluated with the use of Co as an internal and Au as an external monostandard comparators. The concentrations of the elements (Cr, Mn and Zn) in the sediment samples of the present work are discussed regarding to those obtained from other sites.

Thermal neutronflux is determined using the gold wires in current BNCT irradiation, so evaluation of arbitrary points after the irradiation is limited in the quantity of these detectors. In order to make up for the weakness, dose estimation of a patient is simulated by a computational dose calculation supporting system. In another way without computer simulation, a medical irradiation condition can be replicate experimentally using of realistic phantom which was produced from CT images by rapid prototyping technique. This phantom was irradiated at a same JRR-4 neutron beam as clinical irradiation condition of the patient and the thermal neutron distribution on the brain surface was measured in detail. This experimental evaluation technique using a realistic phantom is applicable to in vitro cell irradiation experiments for radiation biological effects as well as in-phantom experiments for dosimetry under the nearly medical irradiation condition of patient. (author)

The EURISOL (The EURopean Isotope Separation On-Line Radioactive Ion Beam) project aims at producing high intensity radioactive ion beams produced by neutron induced fission on a fissile target (235U) surrounding a liquid mercury converter. A proton beam of 1 GeV and 4 MW impinges on the Hg converter generating by spallation reactions high neutronfluxes. In this work the state-of-the-art Monte Carlo codes MCNPX and FLUKA were used to assess the neutronics performance of the system which geometry, inspired from the MAFF concept, allows a versatile manipulation of the fission targets. The objective of the study was to optimize the geometry of the system and the materials used in the fuel and reflector elements of the system, in order to achieve the highest possible fission rate.

Thermal neutronflux is determined using the gold wires in current BNCT irradiation, so evaluation of arbitrary points after the irradiation is limited in the quantity of these detectors. In order to make up for the weakness, dose estimation of a patient is simulated by a computational dose calculation supporting system. In another way without computer simulation, a medical irradiation condition can be replicate experimentally using of realistic phantom which was produced from CT images by rapid prototyping technique. This phantom was irradiated at a same JRR-4 neutron beam as clinical irradiation condition of the patient and the thermal neutron distribution on the brain surface was measured in detail. This experimental evaluation technique using a realistic phantom is applicable to in vitro cell irradiation experiments for radiation biological effects as well as in-phantom experiments for dosimetry under the nearly medical irradiation condition of patient.

For the applicability of instrumental neutron activation analysis (NAA) technique, an irradiation unit with a 37 GBq 241Am-Be neutron source was installed at Institute of Nuclear Sciences of Ankara University. Design and configuration properties of the irradiation unit are described. It has two different sample irradiation positions, one is called site #1 having a pneumatic sample transfer system and the other is site #2 having a location for manual use. In order to characterize neutronflux spectra in the irradiation sites, the measurement results were obtained for thermal (Фth) and epithermal neutronfluxes (Фepi), thermal to epithermal flux ratio (f) and epithermal spectrum shaping factors (α) by employing cadmium ratios of gold (Au) and molybdenum (Mo) monitors. The activities produced in these foils were measured by using a p-type, 44.8% relative efficiency HPGe well detector. For the measured γ-rays, self-absorption and true coincidence summing effects were taken into account. Additionally, thermal neutron self-shielding and resonance neutron self-shielding effects were taken into account in the measured results. For characterization of site #1, the required parameters were found to be Фth = (2.11 ± 0.05) × 103 n cm-2 s-1, Фepi = (3.32 ± 0.17) × 101 n cm-2 s-1, f = 63.6 ± 1.5, α = 0.045 ± 0.009, respectively. Similarly, those parameters were measured in site #2 as Фth = (1.49 ± 0.04) × 103 n cm-2 s-1, Фepi = (2.93 ± 0.15) × 101 n cm-2 s-1, f = 50.9 ± 1.3 and α = 0.038 ± 0.008. The results for f-values indicate that good thermalization of fast neutrons on the order of 98% was achieved in both sample irradiation sites. This is because an optimum combination of water and paraffin moderator is used in the present configuration. In addition, the shielding requirements are met by using natural boron oxide powder (5.5 cm) and boron loaded paraffin layers against neutrons, and a 15 cm thick lead bricks against gamma-rays from source and its

Technological irradiations carried out in material testing reactors (MTRs) are used to study the behavior of materials under irradiation conditions required by different types of nuclear power plants (NPPs). For MTRs, specific instrumentation is required for the experiment monitoring and for the characterization of irradiation conditions, in particular the flux of neutrons and photons. To measure neutron and photon flux in experimental locations, different sensors can be used, such as SPNDs (self-powered neutron detectors), SPGDs (self-powered gamma detectors) and ionization chambers. These sensors involve interactions producing ultimately a measurable electric current. Various sensors have been recently tested in the core periphery of the OSIRIS reactor (located at the CEA-Saclay center) in order to qualify their responses to the neutron and the photon flux. One of the key input data for this qualification is to have a relevant evaluation of neutron and gamma fluxes at the irradiation location. The objective of this work is to evaluate the neutron and the gamma flux in the core periphery of the OSIRIS reactor. With this intention, specific neutron-photonic three-dimensional calculations have been performed and are mainly based on the TRIPOLI-4{sup R} three-dimensional continuous-energy Monte Carlo code, developed by CEA (Saclay Center) and extensively validated against reactor dosimetry benchmarks. In the case of the OSIRIS reactor, TRIPOLI-4{sup R} code has been validated against experimental results based on neutronflux and nuclear heating measurements performed in ex-core and in-core experiments. In this work, simultaneous contribution of neutrons and gamma photons in the core periphery is considered using neutron-photon coupled transport calculations. Contributions of prompt and decay photons have been taken into account for the gamma flux calculation. Specific depletion codes are used upstream to provide the decay-gamma sources required by TRIPOLI-4

The neutronflux at low energy ($E_n\\leq15$ MeV) resulting from the radioactivity of the rock in the underground cavern of the India-based Neutrino Observatory is estimated using Geant4-based Monte Carlo simulations. The neutron production rate due to the spontaneous fission of U, Th and ($\\alpha, n$) interactions in the rock is determined employing the actual rock composition. It has been demonstrated that the total flux is equivalent to a finite size cylindrical rock ($D=L=140$ cm) element. The energy integrated neutronflux thus obtained at the center of the underground tunnel is 2.76 (0.47) $\\times 10^{-6}\\rm~n ~cm^{-2}~s^{-1}$. The estimated neutronflux is of the same order ($\\sim10^{-6}\\rm~n ~cm^{-2}~s^{-1}$)~as measured in other underground laboratories.

Verification of future warhead-dismantlement treaties will require detection of certain warhead attributes without the disclosure of sensitive design information, and this presents an unusual measurement challenge. Neutron spectroscopy—commonly eschewed as an ill-posed inverse problem—may hold special advantages for warhead verification by virtue of its insensitivity to certain neutron-source parameters like plutonium isotopics. In this article, we investigate the usefulness of unfolded neutron spectra obtained from organic-scintillator data for verifying a particular treaty-relevant warhead attribute: the presence of high-explosive and neutron-reflecting materials. Toward this end, several improvements on current unfolding capabilities are demonstrated: deuterated detectors are shown to have superior response-matrix condition to that of standard hydrogen-base scintintillators; a novel data-discretization scheme is proposed which removes important detector nonlinearities; and a technique is described for re-parameterizing the unfolding problem in order to constrain the parameter space of solutions sought, sidestepping the inverse problem altogether. These improvements are demonstrated with trial measurements and verified using accelerator-based time-of-flight calculation of reference spectra. Then, a demonstration is presented in which the elemental compositions of low-Z neutron-attenuating materials are estimated to within 10%. These techniques could have direct application in verifying the presence of high-explosive materials in a neutron-emitting test item, as well as other for treaty verification challenges.

Highlights: • An irradiation unit was installed using a 37 GBq {sup 241}Am-Be neutron source. • The source neutrons moderated by using both water and paraffin. • Irradiation unit was shielded by boron oxide and lead against neutrons and gammas. • There are two sites for irradiations, one of them has a pneumatic transfer system. • Cadmium ratio method was used for irradiation site characterization. - Abstract: For the applicability of instrumental neutron activation analysis (NAA) technique, an irradiation unit with a 37 GBq {sup 241}Am-Be neutron source was installed at Institute of Nuclear Sciences of Ankara University. Design and configuration properties of the irradiation unit are described. It has two different sample irradiation positions, one is called site #1 having a pneumatic sample transfer system and the other is site #2 having a location for manual use. In order to characterize neutronflux spectra in the irradiation sites, the measurement results were obtained for thermal (φ{sub th}) and epithermal neutronfluxes (φ{sub epi}), thermal to epithermal flux ratio (f) and epithermal spectrum shaping factors (α) by employing cadmium ratios of gold (Au) and molybdenum (Mo) monitors. The activities produced in these foils were measured by using a p-type, 44.8% relative efficiency HPGe well detector. For the measured γ-rays, self-absorption and true coincidence summing effects were taken into account. Additionally, thermal neutron self-shielding and resonance neutron self-shielding effects were taken into account in the measured results. For characterization of site #1, the required parameters were found to be φ{sub th} = (2.11 ± 0.05) × 10{sup 3} n cm{sup −2} s{sup −1}, φ{sub epi} = (3.32 ± 0.17) × 10{sup 1} n cm{sup −2} s{sup −1}, f = 63.6 ± 1.5, α = 0.045 ± 0.009, respectively. Similarly, those parameters were measured in site #2 as φ{sub th} = (1.49 ± 0.04) × 10{sup 3} n cm{sup −2} s{sup −1}, φ{sub epi} = (2.93 ± 0

SPHERES is a third-generation neutron backscattering spectrometer, located at the 20 MW German neutron source FRM II and operated by the Jülich Centre for Neutron Science. It offers an energy resolution (fwhm) better than 0.65 μeV, a dynamic range of ± 31 μeV, and a signal-to-noise ratio of up to 1750:1.

SPHERES (SPectrometer with High Energy RESolution) is a third-generation neutron backscattering spectrometer, located at the 20 MW German neutron source FRM II and operated by the J\\"ulich Centre for Neutron Science. It offers an energy resolution (fwhm) better than 0.65 micro-eV, a dynamic range of +-31 micro-eV, and a signal-to-noise ratio of up to 1750:1.

SPHERES is a third-generation neutron backscattering spectrometer, located at the 20 MW German neutron source FRM II and operated by the Juelich Centre for Neutron Science. It offers an energy resolution (fwhm) better than 0.65 {mu}eV, a dynamic range of {+-} 31 {mu}eV, and a signal-to-noise ratio of up to 1750:1.

The optimization processes known as Taguchi method and DOE methodology are applied to the design, training and testing of Artificial Neural Networks in the neutron spectrometry field, which offer potential benefits in the evaluation of the behavior of the net as well as the ability to examine the interaction of the weights and neurons inside the same one. In this work, the Robust Design of Artificial Neural Networks methodology is used to solve the neutron spectra unfolding problem, designing, training and testing an ANN using a set of 187 neutron spectra compiled by the International Atomic Energy Agency, to obtain the better neutron spectra unfolded from the Bonner spheres spectrometer's count rates. (Author)

Hydrogen abundances in lunar polar cold traps are investigated using remotely-sensed neutron count rates. The effect of neutron beaming from craters is measured using data from the Lunar Prospector Neutron Spectrometer (LPNS) and understood in the context of a simple model. This enables a reanalysis of data near the lunar poles, accounting for the topographical impact on the neutron count rates, leading to improved estimates of the hydrogen abundance in the various cold traps. For the case of Cabeus, taking into account the topographical effect increases the inferred water- equivalent hydrogen weight percentage from˜1%to˜4%, consistent with that measured using the LCROSS impactor.

An epithermal neutron beam facility for preclinical neutron capture therapy research has been constructed at the Washington State University TRIGA research reactor installation. Subsequent to a recent upgrade, this new facility offers a high-purity epithermal beam with intensity on the order of 1.2x10(9)n/cm(2)s. Key features include a fluoride-based design for the neutron filtering and moderating components as well as a novel collimator design that allows ease of assembly and disassembly of the beamline components.

For superconducting material to maintain high critical current density, Jc in any applications, effective flux pinning centers are needed. The addition of small size MgO particles in bulk Bi2Sr2CaCu2O8 (Bi-2212) superconductor has been proven to enhance the effective flux pinning centers in the superconducting material by creating a desired microstructure with appropriate defects. To further enhance the pinning properties, radiation is one of the convenient ways to improve the microstructure of the material that has correlation with basic properties of superconductors. Neutron irradiation is one of the niche techniques that can be used to perform the task. Defects with larger radius have dimension comparable to the coherence length of the material and thus improved its superconducting properties. In this paper, a small amount of nanosized MgO particles was used to create defects in the Bi-2212 superconducting material. The Bi-2212/MgO compounds were heat treated, followed by partial melting and slow cooling. Part of the samples was subjected to neutron irradiation using the TRIGA-MARK-II research reactor at the Malaysian Nuclear Agency. Characterization of non-irradiated and irradiated samples was performed via the temperature dependence on electrical resistance measurements, X-ray Diffraction Patterns (XRD), and Scanning Electron Microscopy (SEM) with Energy Dispersive X-ray (EDX) analysis. From the analysis, there was changed in the critical current density and transition temperature of samples subjected to neutron irradiation due to formation of point defects in the microstructure. Higher critical current density indicates better flux pinning properties in the Bi-2212/MgO compounds.

The review describes physical principles underlying efficient production of free neutrons, up-to-date possibilities and prospects of creating fission and fusion neutron sources with intensities of 1015-1021 neutrons/s, and schemes of production and application of neutrons in fusion-fission hybrid systems. The physical processes and parameters of high-temperature plasmas are considered at which optimal conditions for producing the largest number of fusion neutrons in systems with magnetic and inertial plasma confinement are achieved. The proposed plasma methods for neutron production are compared with other methods based on fusion reactions in nonplasma media, fission reactions, spallation, and muon catalysis. At present, intense neutronfluxes are mainly used in nanotechnology, biotechnology, material science, and military and fundamental research. In the near future (10-20 years), it will be possible to apply high-power neutron sources in fusion-fission hybrid systems for producing hydrogen, electric power, and technological heat, as well as for manufacturing synthetic nuclear fuel and closing the nuclear fuel cycle. Neutron sources with intensities approaching 1020 neutrons/s may radically change the structure of power industry and considerably influence the fundamental and applied science and innovation technologies. Along with utilizing the energy produced in fusion reactions, the achievement of such high neutron intensities may stimulate wide application of subcritical fast nuclear reactors controlled by neutron sources. Superpower neutron sources will allow one to solve many problems of neutron diagnostics, monitor nano-and biological objects, and carry out radiation testing and modification of volumetric properties of materials at the industrial level. Such sources will considerably (up to 100 times) improve the accuracy of neutron physics experiments and will provide a better understanding of the structure of matter, including that of the neutron itself.

A nodal formulation of a fixed-source two-dimensional neutron transport problem, in Cartesian geometry, defined in a heterogeneous medium, is solved by an analytical approach. Explicit expressions, in terms of the spatial variables, are derived for averaged fluxes in each region in which the domain is subdivided. The procedure is an extension of an analytical discrete ordinates method, the ADO method, for the solution of the two-dimensional homogeneous medium case. The scheme is developed from the discrete ordinates version of the two-dimensional transport equation along with the level symmetric quadrature scheme. As usual for nodal schemes, relations between the averaged fluxes and the unknown angular fluxes at the contours are introduced as auxiliary equations. Numerical results are in agreement with results available in the literature.

The use of neutron activation foils is a widely spread technique applied to obtain nuclear parameters then comparing the results with those calculated using specific methodologies and available nuclear data. By irradiation of activation foils and subsequent measurement of its induced activity, it is possible to determine the neutronflux at the position of irradiation. The power level during operation of the reactor is a parameter which is directly proportional to the average neutronflux throughout the core. The objective of this work is to gather data from irradiation of gold foils symmetrically placed along a cylindrically configured core which presents only a small excess reactivity in order to derive the power generated throughout the spatial thermal and epithermal neutronflux distribution over the core of the IPEN/MB-01 Nuclear Reactor, eventually lending to a proper calibration of its nuclear channels. The foils are fixed in a Lucite plate then irradiated with and without cadmium sheaths so as to obtain the absolute thermal and epithermal neutronflux. The correlation between the average power neutronflux resulting from the gold foils irradiation, and the average power digitally indicated by the nuclear channel number 6, allows for the calibration of the nuclear channels of the reactor. The reactor power level obtained by thermal neutronflux mapping was (74.65 ± 2.45) watts to a mean counting per seconds of 37881 cps to nuclear channel number 10 a pulse detector, and 0.719.10{sup -5} ampere to nuclear linear channel number 6 (a non-compensated ionization chamber). (author)

Horizontal displacement of plasma along the major radius has been found to significantly influence the fluxes of 2.45 MeV DD neutrons and high-energy charge-exchange atoms from neutral beam injection (NBI) heated plasma of the TUMAN-3M tokamak. An inward shift by Δ R = 1 cm causes 1.2-fold increase in the neutronflux and 1.9-fold increase in the charge-exchange atom flux. The observed increase in the neutronflux is attributed to joint action of several factors-in particular, improved high-energy ion capture and confinement and, probably, decreased impurity inflow from the walls, which leads to an increase in the density of target ions. A considerable increase in the flux of charge-exchange neutrals in inward-shifted plasma is due to the increased number of captured high-energy ions and, to some extent, the increased density of the neutral target. As a result of the increase in the content of high-energy ions, the central ion temperature T i (0) increased from 250 to 350 eV. The dependence of the neutron rate on major radius R 0 should be taken into account when designing compact tokamak-based neutron sources.

Neutron analysis of the first Neutral Beam (NB) for the International Thermonuclear Experimental Reactor (ITER) was performed to provide the basis for the study of the following main aspects: personnel safety during normal operation and maintenance, radiation shielding design, transportability of the NB components in the European countries. The first ITER NB is a medium energy light particle accelerator. In the scenario considered for the calculation the accelerated particles are negative deuterium ions with maximum energy of 1 MeV. The average beam current is 13.3 A. To assess neutron transport in the ITER NB structure a mathematical model of the components geometry was implemented into MCNP computer code (MCNP version 4c2. "Monte Carlo N-Particle Transport Code System." RSICC Computer Code Collection. June 2001). The neutron source definition was outlined considering both D-D and D-T neutron production. FISPACT code (R.A. Forrest, FISPACT-2003. EURATOM/UKAEA Fusion, December 2002) was used to assess neutron...

The following report describes the neutronics results obtained with the MCNP model of the RHF U7Mo LEU reference design that has been established in 2010 during the feasibility analysis. This work constitutes a complete and detailed neutronics analysis of that LEU design using models that have been significantly improved since 2010 and the release of the feasibility report. When possible, the credibility of the neutronics model is tested by comparing the HEU model results with experimental data or other codes calculations results. The results obtained with the LEU model are systematically compared to the HEU model. The changes applied to the neutronics model lead to better comparisons with experimental data or improved the calculation efficiency but do not challenge the conclusion of the feasibility analysis. If the U7Mo fuel is commercially available, not cost prohibitive, a back-end solution is established and if it is possible to manufacture the proposed element, neutronics analyses show that the performance of the reactor would not be challenged by the conversion to LEU fuel.

The Oak Ridge National Laboratory Neutron Sciences Directorate has installed a neutron imaging beamline at the High Flux Isotope Reactor (HFIR) cold guide hall. CG-1D is one of the three instruments that make up the CG1 instrument suite. The beamline optics and detector have recently been upgraded to meet the needs of the neutron imaging community (better "smoothing" of guide system artifacts, higher flux or spatial resolution). These upgrades comprise a new diffuser/aperture system, two new detectors, a He-filled flight tube and silicon (Si) windows. Shielding inside the flight tube, beam scrapers and a beam stop ensure that biological dose is less than 50 μSv/hr outside of the radiation boundary. A set of diffusers and apertures (pinhole geometry) has been installed at the exit of the guide system to allow motorized L/D variation. Samples sit on a translation/rotation stage for alignment and tomography purposes. Detectors for the CG-1D beamline are (1) an ANDOR DW936 charge coupled device (CCD) camera with a field of view of approximately 7 cm x 7 cm and ∼ 80 microns spatial resolution and 1 frame per second time resolution, (2) a new Micro-Channel Plate (MCP) detector with a 2.8 cm x 2.8 cm field of view and 55 microns spatial resolution, and 5 μs timing capability. 6LiF/ZnS scintillators of thickness varying from 50 to 200 microns are being used at this facility. An overview of the beamline upgrade and preliminary data is presented here.

The main goal of this paper is to present a general iteration strategy for solving the discrete form of multidimensional neutron diffusion equations equivalent mathematically to an eigenvalue problem. Usually a solution method is based on different levels of iterations. The presented matrix formalism allows us to visualize explicitly how the used matrix splitting influences the matrix structure in an eigenvalue problem to be solved as well as the interdependence between inner and outer iteration within global iterations. Particular interactive strategies are illustrated by numerical results obtained for several reactor problems. (author). 21 refs, 32 figs, 15 tabs.

The main goal of this paper is to present a general iteration strategy for solving the discrete form of multidimensional neutron diffusion equations equivalent mathematically to an eigenvalue problem. Usually a solution method is based on different levels of iterations. The presented matrix formalism allows us to visualize explicitly how the used matrix splitting influences the matrix structure in an eigenvalue problem to be solved as well as the interdependence between inner and outer iterations within global iterations. Particular iterative strategies are illustrated by numerical results obtained for several reactor problems. (author). 21 refs, 35 figs, 16 tabs.

The FN method which has been applied to many physical problems for isotropic and anisotropic scattering in neutron transport theory is extended for problems for extremely anisotropic scattering. This method depends on the Placzek lemma and the use of the infinite medium Green's function. Here the Green's function for extremely anisotropic scattering which was expressed as a combination of the Green's functions for isotropic scattering is used to solve the critical slab problem. It is shown that the criticality condition is in agreement with the one obtained previously by reducing the transport equation for anisotropic scattering to isotropic scattering and solving using the FN method.

Experiments with the natural uranium target assembly "QUINTA" exposed to 4 and 8 GeV deuteron beams of the Nuclotron accelerator at the Joint Institute for Nuclear Research (Dubna) are analyzed. The reaction rates of 27Al(n,y1)24Na, 27Al(n,y2)22Na and 27Al(n,y3)7Be reactions with effective threshold energies of 5, 27, and 119 MeV were measured at both 4 GeV and 8 GeV deuteron beam energies. The average neutronfluxes between the effective threshold energies and the effective ends of the neutron spectra (which are 800 or 1000 MeV for 4 or 8 GeV deuterons) were determined. The evidence for the intensity shift of the neutron spectra to higher neutron energies with the increase of the deuteron energy from 4 GeV to 8 GeV was found from the ratios of the average neutronfluxes. The reaction rates and the average neutronfluxes were calculated with the MCNPX 2.7 code.

Generally, nuclear flux measurement in research and production reactor are carried out in two stages: first, low level fluxes are measured by counting the pulses produced by fission or boron trifluoride chambers. Second, for high flux levels the parameter measured is the mean current generated in a compensated ionization chamber. A method which shows the feasibility of measuring neutronflux in the second stage with the same counting chamber used in the first stage, without the need to move it from its placement, is presented. (author). 6 refs., 4 figs.

The IPEN/MB-01 reactor, located in the city of Sao Paulo - Brazil, reached its first criticality on the year of 1988. The reactor is characterized by a low output power of 100 W only, even because its purpose is to produce knowledge about nuclear power plants on a smaller geometric scale without the requirement of an extremely complex cooling system. The use of devices such as this it is very interesting because it achieves the demands of nuclear engineering about the neutronic parameters needed in the design of large nuclear plants through relatively simple and inexpensive methods. In this paper, the computational mathematical code MCNP5 is used to perform the calculation of the thermal neutronflux in the core of the IPEN/MB-01 reactor. To do this is used an experiment from the LEU-COMP-THERM-077 benchmark that represents the standard rectangular configuration of the IPEN/MB-01 reactor. The thermal neutronflux is calculated at some axial planes of different heights and, after that, axial profiles of the thermal neutronflux are done and compared to experimental results issued previously. The experimental values used as reference refer to a cylindrical configuration of the core of the reactor. Finally, the pertinence and relevance of the results are checked. With this work is expected to produce more knowledge about the dynamics of neutronflux in the core of the IPEN/MB-01 reactor. (author)

Within the scope of CONRAD (A Coordinated Action for Radiation Dosimetry) Work Package 4 on Computational Dosimetry jointly collaborated with the other research actions on internal dosimetry, complex mixed radiation fields at workplaces and medical staff dosimetry. Besides these collaborative actions, WP4 promoted an international comparison on eight problems with their associated experimental data. A first set of three problems, the results of which are herewith summarised, dealt only with the expression of the stochastic uncertainties of the results: the analysis of the response function of a proton recoil telescope detector, the study of a Bonner sphere neutron spectrometer and the analysis of the neutron spectrum and dosimetric quantity H(p)(10) in a thermal neutron facility operated by IRSN Cadarache (the SIGMA facility). A second paper will summarise the results of the other five problems which dealt with the full uncertainty budget estimate. A third paper will present the results of a comparison on in vivo measurements of the (241)Am bone-seeker nuclide distributed in the knee. All the detailed papers will be presented in the WP4 Final Workshop Proceedings.

Full Text Available In this article we obtain weak solutions for a class nonlinear elliptic problems for the $p(x$-Laplacian-like operators under no-flux boundary conditions. Our result is obtained using a Fredholm-type result for a couple of nonlinear operators, and the theory of variable exponent Sobolev spaces.

Full Text Available Instrumental neutron activation analysis based on the k0 method (k0-INAA requires the availability of the accurate reactor parameter data, in particular a thermal neutronflux that interact with a targets inside the target capsule. This research aims to determine and map the thermal neutronflux inside the capsule and irradiation channels used for the elemental quantification using the k0-AANI. Mapping of the thermal neutronflux (фth on two type of irradiation capsule have been done for RS01 and RS02 facilities of RSG-GAS reactor. Thermal neutronflux determined using Al-0,1%Au alloy through 197Au(n,g 198Au nuclear reaction, while the flux mapping done using statistics R. Thermal neutronflux are calculated using k0-IAEA software provided by IAEA. The results showed the average thermal neutronflux is (5.6±0.3×10+13 n.cm-2.s-1; (5.6±0.4×10+13 n.cm-2.s-1; (5.2±0.4×10+13 n.cm-2.s-1 and (5.3±0.4×10+13 n.cm-2.s-1 for Polyethylene capsule of 1st , 2nd, 3rd and 4th layer respectively. In the case of Aluminum capsule, the thermal neutronflux was lower compared to that on Polyethylene capsule. There were (3.0±0.2×10+13 n.cm-2.s-1; (2.8±0.1×10+13 n.cm-2.s-1; (3.2±0.3×10+13 n.cm-2.s-1 for 1st, 2nd and 3rd layers respectively. For each layer in the capsule, the thermal neutronflux is not uniform and it was no degradation flux in the axial direction, both for polyethylene and aluminum capsules. Contour map of eight layer on polyethylene capsule and six layers on aluminum capsule for RS01 and RS02 irradiation channels had a similar pattern with a small diversity for all type of the irradiation capsule. Keywords: thermal neutron, flux, capsule, NAA Analisis aktivasi neutron instrumental berbasis metode k0 (k0-AANI memerlukan ketersediaan data parameter reaktor yang akurat, khususnya data fluks neutron termal yang berinteraksi dengan inti sasaran di dalam kapsul target. Penelitian ini bertujuan menentukan dan memetakan fluks neutron termal

French requirements in neutron beams of different energies extracted from a reactor are briefly described. The well-known importance of cold neutrons (above 4 Angstrom) is emphasized. The main characteristics of a reactor suitable for physicists are outlined: They are: 1 - A flux of about 7. 10{sup 14} thermal neutrons in the heavy water of the reflector, 2 - Maximum flexibility obtained by: - physical separation of the core and the reflector, - independence of the different experiments, - possibility of modifying physical experiments up to - and including - the nature of the used reflector, without any appreciable interruption in the operation of the reactor, - reduction of fixed shields to a minimum; ample use of liquid shields (water) and fluid shields (sands). 3 - Technological continuity as far as possible with French research reactors (Siloe, Pegase, Osiris) already existing or under construction. 4 - Safety of operation arising from simplicity of conception. 5 - Minimised construction costs. Lowering of the operating costs is looked for indirectly in the simplification of the solutions and the reduction of operating staff, rather than directly by reducing the consumption of fuel elements and energy. The recommended solution can be described as a closed-core non-pressurized swimming-pool reactor, highly under-moderated by the cooling light water. Surrounding the reactor are a number of 'beam tubes-loops' each consisting of: - a part of the reflector (heavy water in the example described), - a part of neutron extraction beam tube, - the circuits required for their cooling, - the inlet systems of suitable fluids to the beam tube nose (liquid hydrogen in the example described), - the necessary outlets for measurement and control system. The whole 'beam tubes loops' is immersed in the water of the metallic self-supporting swimming-pool. The shielding outside the swimming-pool is composed for the most part by heavy sand in which is the rest of

An instrument concept of a reflectometer with a vertical sample geometry fitted to the long pulse structure of a spallation source, called “VERITAS” at the ESS, is presented. It focuses on designing a reflectometer with high intensity at the lowest possible background following the users' demand to investigate thin layers or interfacial areas in the sub-nanometer length scale. The high intensity approach of the vertical reflectometer fits very well to the long pulse structure of the ESS. Its main goal is to deliver as much usable intensity as possible at the sample position and be able to access a reflectivity range of 8 orders of magnitude and more. The concept assures that the reflectivity measurements can be performed in its best way to maximize the flux delivered to the sample. The reflectometer is optimized for studies of (magnetic) layers having thicknesses down to 5Å and a surface area of 1x1cm2. With reflectivity measurements the depth-resolved, laterally averaged chemical and magnetic profile can be investigated. By using polarised neutrons, additional vector information on the in-plane magnetic correlations (off-specular scattering at the pm length scale, GISANS at the nm length scale) can be studied. The full polarisation analysis could be used for soft matter samples to correct for incoherent scattering which is presently limiting neutron reflectivity studies to a reflectivity range on the order of 10-6.

MCNPX computer code, one of the latest versions of code MCNP transport were used to study the flux distribution and its neutronic fluence as a function of energy in two research reactor irradiation IPR-R1. The model developed was validated with research conducted by Dalle (2005). Initially, in the simulation is considered fresh fuel whose core configuration contained three neutron rods control, being two of them 100% ejected while the other inserted 3,1 x 10{sup -1} m deep, as adopted in the literature situation. The neutron source used was the critical type, through KSRC card. The results of the neutron flow and neutronic fluence were obtained in the central tube and the turntable on a range of energy spectrum that ranged from 1.0 x 10{sup -9} MeV to 10 MeV, showing good correlations with the model used in validation. Finally, a hypothetical situation wherein the three reactor control rods are ejected simultaneously was simulated. The simulation results showed an increase in the neutronflux of 7% in the central tube and 5% on the turntable.

This paper describes a study in which only one kind of neutron detector were used in the advanced ENFMS. The conceptual design was performed for overall system with unified fission chamber. The system consists of detector, junction box, wide-range amplifier and signal processing device. Also the requirements of 10CFR50 App. R were considered in design. On the other hand, through computer simulation, the characteristics of pulse-count mode and MSV mode was scrutinized and each noise withstanding capability was analyzed. The results say that 3rd moment has the more stable characteristics to background noise than MSV method. Also, to remain the integrity of information against noise, during installation and operation, the overall system of KSNP was analyzed from a view of noise. By administration for the cause of noise and noise-coupling paths, through the full understanding of noise characteristics, the transfer of the noise source can be minimized. (Author).

In this work was conducted the mapping of the thermal and epithermal neutronsflux and the energy spectrum of the neutrons in the reactor core IPEN/MB-01 for a cylindrical core configuration with minor excess reactivity, which is 28 x 28 fuel rods arranged in north-south and east-west directions. The calibration of control rods for this configuration determined their excess reactivity. The lower excess reactivity in the core decreased neutronflux disturbance caused by the neutron absorbing rods , given that the nuclear reactor was operated with the rods almost completely removed . Was used the 'Activation Analysis Technique' with the thin foil activation detectors ( infinitely diluted and hyper-pure), of different materials that work in different energy ranges, to calculate the saturation activity, used for determining the neutronflux and in the SANDBP code as input for the calculation of the neutrons energy spectrum. To discriminate thermal and epithermal flux , was used the 'Cadmium RatioTechnique' . The activation detectors were distributed in a total of 140 radial and axial positions in the reactor core and 16 irradiation, with bare and covered with cadmium activation foils. A model of this configuration was simulated by MCNP-5 code to determine the cadmium correction factor and comparison of the results obtained experimentally. The cylindrical configuration desired, with 17% less fuel than the standard rectangular configuration (28 x 26 fuel rods), reached criticality with the control rods approximately 90% removed, which decreased considerably the disturbance in neutronflux. Given the highest power density of the 28 x 28 cylindrical core, the neutronflux increased by over 50% in the central regions of the core compared to the values of the 28 x 26 standard rectangular core. (author)

The mathematical solution of the neutron diffusion equation exhibits singularities in its derivatives at material corners. A mathematical treatment of the nature of these singularities and its impact on coarse network approximation methods in computational work is presented. The mathematical behavior is deduced from Green's functions, based on a generalized theory for two space dimensions, and the resulting systems of integral equations, as well as from the Kontorovich--Lebedev Transform. The effect on numerical calculations is demonstrated for finite difference and finite element methods for a two-region corner problem.

In this work, a new mixed discontinuous finite element formulation is applied to the solution of neutron diffusion problems posed in a heterogeneous section of a reactor, taking account of two energy groups. Numerical results of a model problem are presented herein, which demonstrate the efficiency of the method and the little mesh refinement necessary for a good approximation. The above-mentioned formulation has terms that guarantee its stability even without to use of the LBB (Ladyzhesnkaya-Babuska-Brezzi) condition. (author)

In this work is done a technical study for determining Nitrogen (protein) and other elements in wheat flour Activation analysis, with Fast neutrons from a Thermal nuclear reactor. Initially it is given an introduction about the basic principles of the methods of analysis. Equipment used in Activation analysis and a brief description of the neutron source (Thermal nuclear reactor). The realized experiments for determining the flux form in the irradiation site, the half life of N-13 and the interferences due to the sample composition are included too. Finally, the obtained results by Activation and the Kjeldahl method are tabulated. (Author)

The instrumental concept of HERITAGE - a reflectometer with a horizontal sample geometry - well fitted to the long pulse structure of a neutron source is presented. It is constitutes a new class of reflectometers achieving the unprecedentedly high flux for classical specular reflectometry combined with off-specular reflectometry and grazing incidence small-angle scattering (GISANS), thus resulting in a complete 3-d exploration of lateral and in depth structures in thin films. This is achieved by specially designed neutron guides. In the horizontal direction (perpendicular to the scattering plane) the guide's elliptic shape focusses the neutrons onto the sample. In the vertical direction a multichannel geometry provides a smooth divergence distribution at the sample position while accepting the entire beam from a compact high-brilliance flat moderator. The modular collimation setup of HERITAGE provides extremely high flexibility in respect to sample geometries and environments, including the possibility to study virtually all types of solid and liquid interfaces, statically or kinetically. The use of multiple beam illumination allows for reflectivity and GISANS measurements at liquid interfaces both from above and below without a need to move the sample. This concept assures the delivery of the maximum possible and usable flux to the sample in both reflectivity and GISANS measurement regimes. The presented design outperforms the flux of all present-day and already for the ESS planned reflectometers and GISANS setups in flux and in measuring time for standard samples.

The conceptual Fast Energy Amplifier, proposed by Rubbia et al. (1995), consists of a combination of a U-233/Th-232 fuelled fast-neutron subcritical facility with a proton accelerator. An intense beam of 1 GeV protons is injected into liquid lead at the core centre and drives the reactor by producing spallation neutrons. The burst of spallation neutrons produced by a single proton alters the basic neutron statistics which are well known for thermal neutrons in conventional nuclear reactors. A short assessment of standard neutron noise analysis methods is made with respect to monitoring neutron parameter data. (author) 18 refs., 14 figs., 3 tabs.

In this paper, the nonlinear coarse-mesh finite difference method with two-node local problem (CMFD2N) is proven to be unconditionally stable for neutron diffusion eigenvalue problems. The explicit current correction factor (CCF) is derived based on the two-node analytic nodal method (ANM2N), and a Fourier stability analysis is applied to the linearized algorithm. It is shown that the analytic convergence rate obtained by the Fourier analysis compares very well with the numerically measured convergence rate. It is also shown that the theoretical convergence rate is only governed by the converged second harmonic buckling and the mesh size. It is also noted that the convergence rate of the CCF of the CMFD2N algorithm is dependent on the mesh size, but not on the total problem size. This is contrary to expectation for eigenvalue problem. The novel points of this paper are the analytical derivation of the convergence rate of the CMFD2N algorithm for eigenvalue problem, and the convergence analysis based on the analytic derivations.

In this paper, the nonlinear coarse-mesh finite difference method with two-node local problem (CMFD2N) is proven to be unconditionally stable for neutron diffusion eigenvalue problems. The explicit current correction factor (CCF) is derived based on the two-node analytic nodal method (ANM2N), and a Fourier stability analysis is applied to the linearized algorithm. It is shown that the analytic convergence rate obtained by the Fourier analysis compares very well with the numerically measured convergence rate. It is also shown that the theoretical convergence rate is only governed by the converged second harmonic buckling and the mesh size. It is also noted that the convergence rate of the CCF of the CMFD2N algorithm is dependent on the mesh size, but not on the total problem size. This is contrary to expectation for eigenvalue problem. The novel points of this paper are the analytical derivation of the convergence rate of the CMFD2N algorithm for eigenvalue problem, and the convergence analysis based on the analytic derivations.

Monte Carlo method is the powerful technique for solving wide range of radiation transport problems. Its features are that it can solve the Boltzmann`s transport equation almost without approximation, and that the complexity of the systems to be treated rarely becomes a problem. However, the Monte Carlo calculation is always accompanied by statistical errors called variance. In shielding calculation, standard deviation or fractional standard deviation (FSD) is used frequently. The expression of the FSD is shown. Radiation shielding problems are roughly divided into transmission through deep layer and streaming problem. In the streaming problem, the large difference in the weight depending on the history of particles makes the FSD of Monte Carlo calculation worse. The streaming experiment in the 14 MeV neutron rectangular annular bent duct, which is the typical streaming bench mark experiment carried out of the OKTAVIAN of Osaka University, was analyzed by MCNP 4B, and the reduction of variance or FSD was attempted. The experimental system is shown. The analysis model by MCNP 4B, the input data and the results of analysis are reported, and the comparison with the experimental results was examined. (K.I.)

The building of new experimental neutron beam facilities as well as the renewal programmes under development at some of the already existing installations have pinpointed the urgent need to develop neutron guide technology in order to make such neutron transport devices more efficient and durable. In fact, a number of mechanical failures of neutron guides have been reported by several research centres. It is therefore important to understand the behaviour of the glass substrates on top of whi...

Researchers are working on many fronts to make possible high-speed, automated classification and quantification of constituent materials in numerous environments. NASA's Marshall Space Flight Center has implemented a system for rocket engine flowfields/plumes. The Optical Plume Anomaly Detector (OPAD) system was designed to utilize emission and absorption spectroscopy for monitoring molecular and atomic particulates in gas plasma. An accompanying suite of tools and analytical package designed to utilize information collected by OPAD is known as the Engine Diagnostic Filtering System (EDiFiS). The current combination of these systems identifies atomic and molecular species and quantifies mass loss rates in H2/O2 rocket plumes. Capabilities for real-time processing are being advanced on several fronts, including an effort to hardware encode components of the EDiFiS for health monitoring and management. This paper addresses the OPAD with its tool suites, and discusses what is considered a natural progression: a concept for taking OPAD to the next logical level of high energy physics, incorporating fermion and boson particle analyses in measurement of neutronflux.

The instrumental concept of HERITAGE - a reflectometer with a horizontal sample geometry - fitted to the long pulse structure of a neutron source is presented. It is dedicated on creating a new class of reflectometers achieving the unprecedentedly high flux for classical specular reflectometry combined with off-specular reflectometry and grazing incidence small-angle scattering (GISANS), thus resulting in a complete 3-d exploration for lateral structures in thin films. This is achieved by specially designed neutron guides: in the horizontal direction (perpendicular to the scattering plane) it has an elliptic shape and focusses neutrons onto the sample. In the vertical direction it has a multichannel geometry providing a smooth divergence distribution at the sample while accepting the whole beam from a compact high-brilliance flat moderator. The modular collimation setup of HERITAGE provides an extremely high flexibility in respect to sample geometries and environments, including the possibility to study all t...

In the present study, we report a new setup for polarized neutron imaging at the ORNL High Flux Isotope Reactor CG-1D beamline using an in situ 3He polarizer and analyzer. This development is very important for extending the capabilities of the imaging instrument at ORNL providing a polarized beam with a large field-of-view, which can be further used in combination with optical devices like Wolter optics, focusing guides, or other lenses for the development of microscope arrangement. Such a setup can be of advantage for the existing and future imaging beamlines at the pulsed neutron sources. The first proof-of-concept experiment is performed to study the ferromagnetic phase transition in the Fe3Pt sample. We also demonstrate that the polychromatic neutron beam in combination with in situ 3He cells can be used as the initial step for the rapid measurement and qualitative analysis of radiographs.

The complication of critical heat flux (CHF) problem for boiling in channels is caused by the large number of variable factors and the variety of two-phase flows. The existence of several hundreds of correlations for the prediction of CHF demonstrates the unsatisfactory state of this problem. The phenomenological CHF models can provide only the qualitative predictions of CHF primarily in annular-dispersed flow. The CHF look-up tables covered the results of numerous experiments received more recognition in the last 15 years. These tables are based on the statistical averaging of CHF values for each range of pressure, mass flux and quality. The CHF values for regions, where no experimental data is available, are obtained by extrapolation. The correction of these tables to account for the diameter effect is a complicated problem. There are ranges of conditions where the simple correlations cannot produce the reliable results. Therefore, diameter effect on CHF needs additional study. The modification of look-up table data for CHF in tubes to predict CHF in rod bundles must include a method which to take into account the nonuniformity of quality in a rod bundle cross section.

An innovative diamond detector layout is presented that is designed to operate at high temperature under intense neutron and gamma fluxes. It is made of a 500 μm “electronic grade” diamond film with 100 nm thick Ag metal contacts deposited onto each surface of the film by means of thermal evaporation. A 2 μ \\text{m} thick layer of 6LiF has been deposited on top of one of the two Ag contacts to make the detector sensitive to thermal neutrons. The device was tested at the ISIS spallation neutron source (Rutherford Appleton Laboratory, UK) using the INES beam line. The detector was continuously irradiated for 100 hours in vacuum (p = 10-5 \\text{mbar}) , exposed to a neutronflux of about 106 n cm-2 s-1 at a temperature T =150 ^\\circ \\text{C} . The aim of this experiment was to study the time dependence of the diamond detector performance while operating at high temperature under irradiation, providing a first experimental proof of reliable continuous operation for 100 hours at high temperature in a harsh environment.

In order to compensate for the temporary interruption of the PROSPERO reactor, it is desirable to use the CALIBAN reactor in continuous operation. As part of efforts to establish parameters for this service, a study was made of neutronflux and gamma dosage at distances of 50 to 70 cm from the axis of the CALIBAN core. This is a region that has hitherto been little studied during continuous operation. The results are given in this paper.

Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutronfluxes. 7 refs., 22 figs., 11 tabs.

This dissertation has aimed at developing a neutronflux measurement technique by means of detectors activation analysis. The main task of this work was the implementation of this thermal neutronflux measurement technique, using gold wires as activation detectors in the IPEN/MB-01 reactor core. The neutron thermal flux spatial distribution was obtained by gold wire activation technique, with wire diameters of 0.125 mm and 0.250 mm in seven selected reactor experimental channels. The values of thermal flux were about 10{sup 9} neutrons/cm{sup 2}.s. This experiment has been the first one conducted with gold wires in the IPEN/MB-01 reactor, being this technique implemented for use by experiments in flux mapping of the core 73 refs., 60 figs., 31 tabs.

Many techniques for measuring neutron star radii rely on absolute flux measurements in the X-rays. As a result, one of the fundamental uncertainties in these spectroscopic measurements arises from the absolute flux calibrations of the detectors being used. Using the stable X-ray burster, GS 1826-238, and its simultaneous observations by Chandra HETG/ACIS-S and RXTE/PCA as well as by XMM-Newton EPIC-pn and RXTE/PCA, we quantify the degree of uncertainty in the flux calibration by assessing the differences between the measured fluxes during bursts. We find that the RXTE/PCA and the Chandra gratings measurements agree with each other within their formal uncertainties, increasing our confidence in these flux measurements. In contrast, XMM-Newton EPIC-pn measures 14.0$\\pm$0.3% less flux than the RXTE/PCA. This is consistent with the previously reported discrepancy with the flux measurements of EPIC-pn, compared to EPIC-MOS1, MOS2 and ACIS-S detectors. We also address the calibration uncertainty in the RXTE/PCA int...

This paper compiles an experience in Asco I Nuclear Power Plant about electromagnetic interferences associated to the neutronflux measurement system, Source Range Asco I NPP. The circuit affected is the proportional detector (BF3) located outside the reactor vessel to measure the neutron leakage in shutdown and in start-up.

The neutronic simulation of a nuclear reactor core is performed using the neutron transport equation, and leads to an eigenvalue problem in the steady-state case. Among the deterministic resolution methods, diffusion approximation is often used. For this problem, the MINOS solver based on a mixed dual finite element method has shown his efficiency. In order to take advantage of parallel computers, and to reduce the computing time and the local memory requirement, we propose in this dissertation two domain decomposition methods for the resolution of the mixed dual form of the eigenvalue neutron diffusion problem. The first approach is a component mode synthesis method on overlapping sub-domains. Several Eigenmodes solutions of a local problem solved by MINOS on each sub-domain are taken as basis functions used for the resolution of the global problem on the whole domain. The second approach is a modified iterative Schwarz algorithm based on non-overlapping domain decomposition with Robin interface conditions. At each iteration, the problem is solved on each sub domain by MINOS with the interface conditions deduced from the solutions on the adjacent sub-domains at the previous iteration. The iterations allow the simultaneous convergence of the domain decomposition and the eigenvalue problem. We demonstrate the accuracy and the efficiency in parallel of these two methods with numerical results for the diffusion model on realistic 2- and 3-dimensional cores. (author)

The Joint Research Centre has taken into operation a new experimental device designed for research in the fields of nuclear safeguards and security applications. The research projects currently undertaken include detection of shielded contraband materials, detection of fissile materials, and mass determination of small fissile materials in shielded containers. The device, called the Pulsed Neutron Interrogation Test Assembly (PUNITA), incorporates a pulsed 14‐MeV (D‐T) neutron generator and a...

The safety and optimal performance of large, commercial, light-water reactors require the knowledge at all time of the neutron-flux distribution in the core. In principle, this information can be obtained by solving the time-dependent neutron diffusion equations. However, this approach is complicated and very expensive. Sufficiently accurate, real-time calculations (time scale of approximately one second) are not yet possible on desktop computers, even with fast-running, nodal kinetics codes. A semi-experimental, nodal synthesis method which avoids the solution of the time-dependent, neutron diffusion equations is described. The essential idea of this method is to approximate instantaneous nodal group-fluxes by a linear combination of K, precomputed, three-dimensional, static expansion-functions. The time-dependent coefficients of the combination are found from the requirement that the reconstructed flux-distribution agree in a least-squares sense with the readings of J ({ge}K) fixed, prompt-responding neutron-detectors. Possible numerical difficulties with the least-squares solution of the ill-conditioned, J-by-K system of equations are brought under complete control by the use of a singular-value-decomposition technique. This procedure amounts to the rearrangement of the original, linear combination of K expansion functions into an equivalent more convenient, linear combination of R ({le}K) orthogonalized modes'' of decreasing magnitude. Exceedingly small modes are zeroed to eliminate any risk of roundoff-error amplification, and to assure consistency with the limited accuracy of the data. Additional modes are zeroed when it is desirable to limit the sensitivity of the results to measurement noise.

The safety and optimal performance of large, commercial, light-water reactors require the knowledge at all time of the neutron-flux distribution in the core. In principle, this information can be obtained by solving the time-dependent neutron diffusion equations. However, this approach is complicated and very expensive. Sufficiently accurate, real-time calculations (time scale of approximately one second) are not yet possible on desktop computers, even with fast-running, nodal kinetics codes. A semi-experimental, nodal synthesis method which avoids the solution of the time-dependent, neutron diffusion equations is described. The essential idea of this method is to approximate instantaneous nodal group-fluxes by a linear combination of K, precomputed, three-dimensional, static expansion-functions. The time-dependent coefficients of the combination are found from the requirement that the reconstructed flux-distribution agree in a least-squares sense with the readings of J (≥K) fixed, prompt-responding neutron-detectors. Possible numerical difficulties with the least-squares solution of the ill-conditioned, J-by-K system of equations are brought under complete control by the use of a singular-value-decomposition technique. This procedure amounts to the rearrangement of the original, linear combination of K expansion functions into an equivalent more convenient, linear combination of R (≤K) orthogonalized ``modes`` of decreasing magnitude. Exceedingly small modes are zeroed to eliminate any risk of roundoff-error amplification, and to assure consistency with the limited accuracy of the data. Additional modes are zeroed when it is desirable to limit the sensitivity of the results to measurement noise.

The typical fast neutron detector falls into one of two categories, Bonner sphere spectrometers and liquid scintillator proton recoil detectors. These two detector types have traditionally been used to measure fast neutrons at the surface and in low background environments. The cosmogenic neutron spectrum and flux is an important parameter for a number of experimental efforts, including procurement of low background materials and the prediction of electronic device faults. Fast neutrons can also cause problems for underground low-background experiments, through material activation or signals that mimic rare events. Current detector technology is not sufficient to properly characterize these backgrounds. To this end, the University of Maryland and the National Institute of Standards and Technology designed, developed, and deployed two Fast Neutron Spectrometers (FaNS) comprised of plastic scintillator and 3He proportional counters. The detectors are based upon capture-gated spectroscopy, a technique that demands a delayed coincidence between a neutron scatter and the resulting neutron capture after thermalization. This technique provides both particle identification and knowledge that the detected neutron fully thermalized. This improves background rejection capabilities and energy resolution. Presented are the design, development, and deployment of FaNS-1 and FaNS-2. Both detectors were characterized using standard fields at NIST, including calibrated 252Cf neutron sources and two monoenergetic neutron generators. Measurements of the surface fast neutron spectrum and flux have been made with both detectors, which are compared with previous measurements by traditional detectors. Additionally, FaNS-1 was deployed at the Kimballton Underground Research Facility (KURF) in Ripplemead, VA. A measurement of the fast neutron spectrum and flux at KURF is presented as well. FaNS-2 is currently installed in a shallow underground laboratory where it is measuring the muon

Based on the Miniature Neutron Source Reactor (MNSR) with high enrichment uranium (HEU) fuel and accordance with the requirements of BNCT, the 19.5% of enriched fuel UO2 fuel core for BNCT with epithermal neutron treatment beam was primary designed, the reactor core parameters such as epithermal neutronflux density,epithermal neutronflux unit of fast neutron dose rate,epithermal neutronflux unit photon dose rate of γ,epithermal neutronflux ratio of thermal neutronflux, neutron spectrum were calculated and analyzed. The results show that the design program was an excellent epithermal neutron treatment beam.%根据硼中子俘获治疗( BNCT)中子源的要求,在高浓铀为燃料的微型反应堆(MNSR)的基础上,以富集度19.5％的UO2为燃料,将其堆芯低浓化并且添加水平超热中子束流治疗孔道,开展超热中子束流BNCT堆堆芯低浓化初步设计.计算BNCT堆的超热中子注量率、单位超热中子注量的快中子剂量率、单位超热中子注量的γ光子剂量率、超热中子注量与热中子的注量之比、中子束流能谱等关键参数.结果表明,该设计可以得到优良的超热中子束流.

Full Text Available The flux expansion nodal method is a suitable method for considering nodalization effects in node corners. In this paper we used this method to solve the intra-nodal flux analytically. Then, a computer code, named MA.CODE, was developed using the C# programming language. The code is capable of reactor core calculations for hexagonal geometries in two energy groups and three dimensions. The MA.CODE imports two group constants from the WIMS code and calculates the effective multiplication factor, thermal and fast neutronflux in three dimensions, power density, reactivity, and the power peaking factor of each fuel assembly. Some of the code's merits are low calculation time and a user friendly interface. MA.CODE results showed good agreement with IAEA benchmarks, i. e. AER-FCM-101 and AER-FCM-001.

Silicon pillar structures filled with a neutron converter material (10B) are designed to have high thermal neutron detection efficiency with specific dimensions of 50 μm pillar height, 2 μm pillar diameter and 2 μm spacing between adjacent pillars. In this paper, we have demonstrated such a detector has a high neutron-to-gamma discrimination of 106 with a high thermal neutron detection efficiency of 39% when exposed to a high gamma-ray field of 109 photons/cm2s.

Silicon pillar structures filled with a neutron converter material (10B) are designed to have high thermal neutron detection efficiency with specific dimensions of 50 μm pillar height, 2 μm pillar diameter and 2 μm spacing between adjacent pillars. In this paper, we have demonstrated such a detector has a high neutron-to-gamma discrimination of 106 with a high thermal neutron detection efficiency of 39% when exposed to a high gamma-ray field of 109 photons/cm2s.

The three-dimensional polycinetic Monte Carlo particle transport code TRIPOLI has been under development in the French Shielding Laboratory at Saclay since 1965. TRIPOLI-1 began to run in 1970 and became TRIPOLI-2 in 1978: since then its capabilities have been improved and many studies have been performed. TRIPOLI can treat stationary or time dependent problems in shielding and in neutronics. Some examples of solved problems are presented to demonstrate the many possibilities of the system. (author).

This report addresses the solution of Riemann problems for hyperbolic equations when the nonlinear characteristic fields loose their genuine nonlinearity. In this context, exact solvers for nonconvex 1D Riemann problems are developed. First a scalar conservation law for a nonconvex flux with two inflection points is studied. Then the P-system for an isothermal version of the van der Waals gas model is examined in a range of temperatures allowing for a nonconvex pressure function. Eventually the system of the Euler equations of gasdynamics is considered for the polytropic van der Waals gas. In this case, a suitably large specific heat is considered such that the isentropes display a local loss of convexity near the saturation curve and the critical point. Such a nonconvex physical model allows for nonclassical waves to appear as a result of the change of sign of the fundamental derivative of gasdynamics. The solution of the Riemann problem for the considered real gas model reduces to a system of two nonlinear ...

Neutron fields generated in extended heavy (Z\\geq 82) targets under irradiation with proton beams at energies in the range of 1 GeV are investigated. Influence of different moderators on the spectra and multiplicities of neutrons escaping the surface of the assembly consisting of a lead target (\\varnothing 8 cm\\times 20 cm or \\varnothing 8cm\\times 50 cm) screened by variable thickness of polyethylene or graphite, respectively, was compared in the present work. It is shown that the effectiveness of graphite as a material used in such assemblies to moderate spallation neutrons down to thermal energies is significantly lower than that of paraffin.

France has a long experience of about 50 years in designing, building and operating sodium-cooled fast reactors (SFR) such as RAPSODIE, PHENIX and SUPER PHENIX. Fast reactors feature the double capability of reducing nuclear waste and saving nuclear energy resources by burning actinides. Since this reactor type is one of those selected by the Generation IV International Forum, the French government asked, in the year 2006, CEA, namely the French Alternative Energies and Atomic Energy Commission, to lead the development of an innovative GEN-IV nuclear- fission power demonstrator. The major objective is to improve the safety and availability of an SFR. The neutronflux monitoring (NFM) system of any reactor must, in any situation, permit both reactivity control and power level monitoring from startup to full power. It also has to monitor possible changes in neutronflux distribution within the core region in order to prevent any local melting accident. The neutron detectors will have to be installed inside the reactor vessel because locations outside the vessel will suffer from severe disadvantages; radially the neutron shield that is also contained in the reactor vessel will cause unacceptable losses in neutronflux; below the core the presence of a core-catcher prevents from inserting neutron guides; and above the core the distance is too large to obtain decent neutron signals outside the vessel. Another important point is to limit the number of detectors placed in the vessel in order to alleviate their installation into the vessel. In this paper, we show that the architecture of the NFM system will rely on high-temperature fission chambers (HTFC) featuring wide-range flux monitoring capability. The definition of such a system is presented and the justifications of technological options are brought with the use of simulation and experimental results. Firstly, neutron-transport calculations allow us to propose two in-vessel regions, namely the above-core and under

A new deterministic-Monte Carlo hybrid solution technique is derived for the time-dependent transport equation. This new approach is based on dividing the time domain into a number of coarse intervals and expanding the transport solution in a series of polynomials within each interval. The solutions within each interval can be represented in terms of arbitrary source terms by using precomputed response functions. In the current work, the time-dependent response function computations are performed using the Monte Carlo method, while the global time-step march is performed deterministically. This work extends previous work by coupling the time-dependent expansions to space- and angle-dependent expansions to fully characterize the 1D transport response/solution. More generally, this approach represents and incremental extension of the steady-state coarse-mesh transport method that is based on global-local decompositions of large neutron transport problems. An example of a homogeneous slab is discussed as an example of the new developments.

Net radiation provides most of the melt energy for seasonal snow, a critical water resource for many parts of the world. In many cases shortwave radiation is the dominant flux, but when it is reduced by factors such as high albedo, cloud cover, and topographic shading, longwave radiation can also contribute substantially to the surface energy balance. Methods for determining these surface fluxes include: numerical weather models, reanalysis, direct observations, satellite measurements, and empirical algorithms based on proxy data. Long- and shortwave irradiances are rarely measured in mountainous environments. Those measurements that are made in these locations are subject to difficult conditions, which often result in snow-covered instruments and tilted instrumentation or sloped installation surface. To avoid these problems, measurements may be taken from a more protected valley location, but this may lead to a mismatch between measurement and study site conditions, such as a fog covered valley observation used to force a simulation at a higher, fog-free elevation. Satellites are useful tools for observing surface fluxes over large areas. However, satellite data products can have problems with mixed pixels of clouds and no clouds. Finally, algorithms based on proxy data have known biases and errors, can lack cloud and topographic corrections, and may not represent the diurnal cycle or cloud cover variability. In this study, we explored four scenarios for estimating long- and shortwave surface irradiances that have known errors and assessed the impact of these errors on simulations of SWE. The four scenarios were: 1) improper instrument siting such as a tilted sensor or improper projection of observations onto sloped terrain, 2) biases and errors in surface irradiances characteristic of algorithms based on proxy data, 3) mixed pixels of cloudy and non-cloudy conditions resulting from a coarse spatial resolution in a satellite or reanalysis product, and 4) lack of

One of the main objectives related to the safety of any nuclear power plant, including the nuclear power plant of Laguna Verde is to ensure the structural integrity of reactor pressure vessel. To identify and quantify the damage caused by neutron irradiation in the vessel of any nuclear reactor, it is necessary to know both the neutronflux and the neutron fluence that the vessel has been receiving during its operation lifetime, and that the damage observed by mechanical testing are products of microstructural effects induced by neutron irradiation; therefore, it is important the study and prediction of the neutronflux in order to have a better understanding of the damage that these materials are receiving. The calculation here described uses the DORT code, which solves the neutron transport equation in discrete ordinates in two dimensions (x-y, r-{theta} and r-z), according to a regulatory guide, it should make an approximation of the neutronflux in three dimensions by the so called synthesis method. It is called in that way because it achieves a representation of 3 Dimensional neutronflux combining or summarizing the fluxes calculated by DORT r-{theta}, r-z and r. This work presents the application of synthesis method, according to Regulatory Guide 1190, to determine the 3 Dimensional fluxes in internal BWR reactor using three different spatial meshes. The results of the neutronflux and fluence, using three different meshes in the directions r, {theta} and z were compared with results reported in the literature obtaining a difference not larger than 9.61%, neutronflux reached its maximum, 1.58 E + 12 n/cm{sup 2}s, at a height H 4 (239.07 cm) and angle 32.236{sup o} in the core shroud and 4.00 E + 09 n/cm{sup 2}s at a height H 4 and angle 35.27{sup o} in the inner wall of the reactor vessel, positions that are consistent to within {+-}10% over the ones reported in the literature. (Author)

This thesis comes within the framework of minor-actinide nuclear transmutation studies. First of all, we have evaluated the impact of minor actinide nuclear data uncertainties within the cases of {sup 241}Am and {sup 237}Np incineration in three different reactor spectra: EFR (fast), GT-MHR (epithermal) and HI-HWR (thermal). The nuclear parameters which give the highest uncertainties were thus highlighted. As a result of fact, we have tried to reduce data uncertainties, in the thermal energy region, for one part of them through experimental campaigns in the moderated high intensity neutronfluxes of ILL reactor (Grenoble). These measurements were focused onto the incineration and transmutation of the americium-241, the curium-244 and the californium-249 isotopes. Finally, the values of 12 different cross sections and the {sup 241}Am isomeric branching ratio were precisely measured at thermal energy point. (author)

The physical model of the 300 # swimming pool research reactor(SPRR-300) based on the Monte Carlo code MCNP has been verified. Sophisticated modeling is conducted. An effective multiplication factor value of 1. 002 29 is obtained, existing a relative error of 0. 229% compared with the critical value. Meanwhile, a problem comes out that the interrupt and con-tinue-run with parallel version of MCNP doesn't work. The problem is solved through trail and error process. A reasonable application of flux tally average over a cell and flux tally at a point is suggested, namely the former is prior to the latter to tally in big volume. Comparison between calculation results and experimental data shows that the thermal neutronflux has a deviation of 4. 6% at a power level of 3 MW. That is to say, the calculated value and the experimental value agree well with each other, and the neutronflux result is dependable.%基于MCNP程序对300#研究堆首炉堆芯进行精细建模,通过并行计算方式得到了实验临界棒位下堆芯的有效增殖因数为1.002 29,与临界值之间的相对误差为0.229％,验证了物理模型的正确性.探讨并解决了并行计算的中断与接续问题,提出了体通量计数与点探测器计数应用中的合理化建议,即对大体积空间计数时尽量使用体通量计数.计算值与实验值对比结果表明:两者在3 MW功率水平下热中子通量密度相差4.6％,符合得较好.

A characteristics transport theory code, CHAPLET, has been developed for the purpose of making it practical to perform a whole LWR core calculation with the same level of calculational model and accuracy as that of an ordinary single assembly calculation. The characteristics routine employs the CACTUS algorithm for drawing ray tracing lines, which assists the two key features of the flux solution in the CHAPLET code. One is the direct neutron path linking (DNPL) technique which strictly connects angular fluxes at each assembly interface in the flux solution separated between assemblies. Another is to reduce the required memory storage by sharing the data related to ray tracing among assemblies with the same configuration. For faster computation, the coarse mesh rebalance (CMR) method and the Aitken method were incorporated in the code and the combined use of both methods showed the most promising acceleration performance among the trials. In addition, the parallelization of the flux solution was attempted, resulting in a significant reduction in the wall-clock time of the calculation. By all these efforts, coupled with the results of many verification studies, a whole LWR core heterogeneous transport theory calculation finally became practical. CHAPLET is thought to be a useful tool which can produce the reference solutions for analyses of an LWR (author)

中子通量监测器(NFM)可实现ITER实时的中子通量测定,转化得到聚变功率,功率密度,等离子体温度等.获得NFM探测效率对能量的相对平坦响应对准确诊断十分必要.论文针对特定的NFM裂变室结构,运用MCNP—4C对裂变室包裹层慢化剂/屏蔽材料种类及厚度进行了优化计算.这些工作对探测器裂变室结构的优化设计实验标定及定型具有重要意义.%The NeutronFlux Monitor(NFM) can provide the real - time flux of ITER( International Thermonuclear Experimental Reactor ) , and get the fusion power and temperature of the plasma after transformation. A relative flat energy response curve of neutron detection efficiency is essential for accurate diagnosis of NFM in ITER. The paper makes an optimal computation on thickness of different moderator/ shielding material with the MCNP - 4C as to the specific structure of NFM fission chamber. It is significant for the optimal design and the experimental calibration of the NFM

Particles count rates at given Earth location and altitude result from the convolution of (i) the interstellar (IS) cosmic-ray fluxes outside the solar cavity, (ii) the time-dependent modulation of IS into Top-of-Atmosphere (TOA) fluxes, (iii) the rigidity cut-off (or geomagnetic transmission function) and grammage at the counter location, (iv) the atmosphere response to incoming TOA cosmic rays (shower development), and (v) the counter response to the various particles/energies in the shower. Count rates from neutron monitors or muon counters are therefore a proxy to solar activity. In this paper, we review all ingredients, discuss how their uncertainties impact count rate calculations, and how they translate into variation/uncertainties on the level of solar modulation $\\phi$ (in the simple Force-Field approximation). The main uncertainty for neutron monitors is related to the yield function. However, many other effects have a significant impact, at the 5-10% level on $\\phi$ values. We find no clear ranking...

Glioblastoma multiforme is the most common and aggressive of brain tumors and is difficult to treat by surgery, chemotherapy or conventional radiation therapy. One treatment alternative is the Neutron Capture Therapy in Boron, which requires a beam modulated in neutron energy and a drug with {sup 10}B able to be fixed in the tumor. When the patients head is exposed to the neutron beam, they are captured by the {sup 10}B and produce a nucleus of {sup 7}Li and an alpha particle whose energy is deposited in the cancer cells causing it to be destroyed without damaging the normal tissue. One of the problems associated with this therapy is to have an epithermal neutronsflux of the order of 10{sup 9} n/cm{sup 2}-sec, whereby irradiation channels of a nuclear research reactor are used. In this work using Monte Carlo methods, the neutron spectra obtained in the radial irradiation channel of the TRIGA Mark III reactor are calculated when inserting filters whose position and thickness have been modified. From the arrangements studied, we found that the Fe-Cd-Al-Cd polyethylene filter yielded a ratio between thermal and epithermal neutronfluxes of 0.006 that exceeded the recommended value (<0.05), and the dose due to the capture gamma rays is lower than the dose obtained with the other arrangements studied. (Author)

Knowing the magnitude of the neutronflux in the reactor irradiation facilities, is so much importance for the operation of the same one, like for the investigation developing. Particularly, knowing with certain precision the spectrum and the neutronflux in the different positions of irradiation of a reactor, it is essential for the evaluation of the results obtained for a certain irradiation experiment. The TRIGA Mark III reactor account with irradiation facilities designed to carry out experimentation, where the reactor is used like an intense neutron source and gamma radiation, what allows to make irradiations of samples or equipment in radiation fields with components and diverse levels in the different facilities, one of these irradiation facilities is the Thermal Column where the Hohlraum is. In this work it was carried out a characterization of the neutronflux inside the 'Hohlraum' of the irradiation facility Thermal Column of the TRIGA Mark III reactor of the Nuclear Center of Mexico to 1 MW of power. It was determined the sub cadmic neutronflux and the epi cadmic by means of the neutron activation technique of thin sheets of gold. The maps of the distribution of the neutronflux for both energy groups in three different positions inside the 'Hohlraum' are presented, these maps were obtained by means of the irradiation of undressed thin activation sheets of gold and covered with cadmium in arrangements of 10 x 12, located parallel to 11.5 cm, 40.5 cm and 70.5 cm to the internal wall of graphite of the installation in inverse address to the position of the reactor core. Starting from the obtained values of neutronflux it was found that, for the same position of the surface of irradiation of the experimental arrangement, the relative differences among the values of neutronflux can be of 80%, and that the differences among different positions of the irradiation surfaces can vary until in a one order of magnitude. (Author)

Highlights: • The TRIGA IPR-R1 was modelled using MCNP. • The thermal neutronflux through the samples in eleven irradiation channels was obtained. • The simulated results were compared to experimental values. • The relative error, the relative trend, the z-score test and uncertainty were analysed. - Abstract: The TRIGA IPR-R1 was modelled using MCNP. The model consists of a cylinder filled with water, fuel elements, radial reflectors, central tube, control rods and neutron source. Around the core is placed the Rotary Specimen Rack (RSR) with adequate groove to insert the samples to irradiation. The values of the thermal neutronflux through the samples in eleven irradiation channels were simulated and compared to the experimental results to validate the model. After that, the values of the thermal neutronflux, in the same channels, were simulated on two horizontal planes at different heights and compared to validate the model. These channels were characterized as representative channels of the neutronflux distribution in the RSR. To evaluate the results, the relative errors, the relative trend, the z-score test and the relevance to a confidence interval of 95% were analysed. Good agreement has been obtained for the most channels when compared with the experimental results.

The characterization of the fast neutron fluence has become a critical issue for experiments that require extreme low-background environments, such as neutrino-less double-beta decay, dark matter searches, and solar neutrino experiments. In such experiments, fast neutrons may be the dominant and a potentially irreducible background, thus necessitating precise information about the fast neutron fluence and energy spectrum. The most reasonable approach to addressing the problem is through the complete characterization of the neutrons through both site-specific measurement and benchmarking of simulation codes. We will discuss the progress toward the development of a large-volume, segmented detector consisting of plastic scintillator and ^3He proportional counters. The detector will be placed in an underground environment to measure the fast neutronflux and energy spectrum. A prototype detector has been constructed and testing is in progress. We will discuss the status of the project and present data from the prototype detector. )

Full Text Available Dosimetry measurements were carried out during the AMMON experimental program, in the EOLE facility. Al-0.1 wt% Au wires were positioned along curved fuel plates of JHR-type assemblies to investigate the azimuthal and axial gold capture rate profiles, directly linked to the thermal and epithermal flux. After irradiation, wires were cut into small segments (a few mm, and the gold capture rate of each part was measured by gamma spectrometry on the MADERE platform. This paper presents results in the “hafnium” configuration, and more specifically the azimuthal flux profile characterization. The final uncertainty on each measured wire lies below 1% (at 2 standard deviations. Experimental profiles are in a good agreement against Monte Carlo calculations, and the 4% capture rate increase at the plate edge is well observed. The flux dissymmetry due to assembly position in the core is also measured, and shows a 10% discrepancy between the two edges of the plate.

Since the appearance of high fluxneutron beams, scientists experimented with neutron radiography. This high beam flux combined with modern neutron to visible light converters leads to the possibility of performing fast neutron micro-tomography. The first results of cold neutron tomography with a neutron lens are presented in this article. Samples are rotated in the beam and the projections are recorded with a neutron camera. The 3D reconstruction is performed with cone beam reconstruction software.

Neutron small-angle diffraction has been used to investigate the flux-line lattice structure within single crystals of the high-temperature superconductor Bi2.15Sr1.95CaCu2O8+x. The diffracted intensity goes rapidly to zero as the magnetic field or the temperature is increased. Melting at low...

核测量系统采用标准输出信号后,采用自动换档电流放大器的功率测量装置在档位切换时,输出信号有效值与量程的同步性可能出现问题,使保护系统、功率控制系统和报警系统信号接收端产生误动作.本工作对分析接口问题产生的原因进行分析,并提出解决办法.%When nuclear reactor control and protection system uses a unified standard electrical analog signal(4~20 mA)as an exchange signal, there exist problems of synchronization between the virtual value and range of output signals during switching of NeutronFlux Measurement System (NFMS) , and thus false actions occur in the protection system, the alarm system and the power control system. The cause of these new interface issues were described in this paper, and one solution for this problem was given in detail.

The peripatetic ORNL HFIR Center for Neutron Scattering reflectometer instrument MIRROR has recently been re-installed in an interim beam line position in the reactor beam room. In 2006 an upgraded version of the instrument will move to a high intensity guide hall position fed by the new HFIR cold source. In this short note, we present some aspects of current instrument operation-particularly with respect to data reduction from the instrument's linear reflection plane detector-with examples of ongoing research and analysis, and a brief outline of the expected capabilities of the fully upgraded guide hall instrument.

The beginning of a new era of nuclear reactor requires technological advances and also multiples studies. The European Liquid metal cooled Fast breeder Reactor is one of the designs for the generation IV nuclear reactor, selected by ENEA. A pioneer of its time, ELFR needs a demonstrator in order to prove the feasibility of this project and to acquire more data and experience in operating a LFR. For this reason the ALFRED project was started and it is expected to be under operation by the year 2030. This paper has the objective of analyzing the neutronic and thermohydraulics of the ALFRED core by the means of a coupled scheme. The selected code for neutronic simulation is MCNP and the selected code for thermohydraulics is ANSYS.

The Large Core Code Evaluation Working Group Benchmark Problem Four was specified to provide a stringent test of the current methods which are used in the nuclear design and analyses process. The benchmark specifications provided a base for performing detailed burnup calculations over the first two irradiation cycles for a large heterogeneous fast reactor. Particular emphasis was placed on the techniques for modeling the three-dimensional benchmark geometry, and sensitivity studies were carried out to determine the performance parameter sensitivities to changes in the neutronics and burnup specifications. The results of the Benchmark Four calculations indicated that a linked RZ-XY (Hex) two-dimensional representation of the benchmark model geometry can be used to predict mass balance data, power distributions, regionwise fuel exposure data and burnup reactivities with good accuracy when compared with the results of direct three-dimensional computations. Most of the small differences in the results of the benchmark analyses by the different participants were attributed to ambiguities in carrying out the regionwise flux renormalization calculations throughout the burnup step.

A method for combined 3D/1D-modeling of thermohydraulics of a once-through steam generator (SG) based on the joint analysis of three-dimensional thermo- and hydrodynamics of a single-phase heating coolant in the intertube space and one-dimensional thermohydraulics of steam-generating channels (tubes) with the use of well-known friction and heat-transfer correlations under various boiling conditions is discussed. This method allows one to determine the spatial distribution of temperatures and heat fluxes of heat-exchange surfaces of SGs with a single-phase heating coolant in the intertube space and with steam generation within tubes. The method was applied in the analytical investigation of typical operation of a once-through SG of a nuclear power installation with an RBEC fast-neutron heavy-metal reactor that is being designed by Kurchatov Institute in collaboration with OKB GIDROPRESS and Leipunsky Institute of Physics and Power Engineering. Flow pattern and temperature fields were obtained for the heavy-metal heating coolant in the intertube space. Nonuniformities of heating of the steam-water coolant in different heat-exchange tubes and nonuniformities in the distribution of heat fluxes at SG heat-exchange surfaces were revealed.

Fast radio bursts (FRBs) are an emerging class of short and bright radio transients whose sources remain enigmatic. Within the galactic center, the non-detection of pulsars within the inner $\\sim \\!10\\,{\\rm pc}$ has created a missing pulsar problem that has intensified with time. With all reserve, we advance the notion that the two problems could be linked by a common solution: the collapse of neutron stars (NS) due to capture and sedimentation of dark matter (DM) within their cores. Bramante \\& Linden (2014), Phys.\\ Rev.\\ Lett.~19, 191301 showed that certain DM properties allow for rapid NS collapse within the high DM density environments near galactic centers while permitting NS survival elsewhere. Each DM-induced collapse could generate an FRB as the NS magnetosphere is suddenly expelled. This scenario could explain several features of FRBs: their short time scales, large energies, locally produced scattering tails, and high event rates. Our scenario predicts that FRBs are localized to galactic centers...

提出了在强γ环境中脉冲中子通量的薄膜闪烁体测量方法. 根据其与中子、γ响应的理论计算结果,研制成功一种对γ不灵敏,用于探测快脉冲中子通量的新型探测器. 该探测器由塑料薄膜闪烁体+光电探测器构成. 与传统探测器相比,该探测器具有如下特点:1.高中子灵敏度;2.高n/γ分辨;3.在给定能区具有平坦的能量响应.%A new conception of measuring pulsed neutronflux in high gamma-ray environment with a thin plastic scintillating foil is presented. Based on the calculations of the response to neutrons and gamma-rays, a new gamma-insensitive detector for detecting fast rising, transient neutronflux has been developed and preliminarily tested, which comprises a thin plastic scintillating foil of ST401 (TPSF) and a photomultiplier tube (or a photodiode). The detector exhibits three distinct properties compared with the conventional ones: (1) high neutron sensitivity, (2) high n/γ discrimination, and (3) flat response in the given neutron energy range.

Full Text Available The OECD UAM Benchmark was launched in 2005 with the objective of determining the uncertainty in the simulation of Light Water Reactors (LWRs system calculations at all the stages of the coupled reactor physics—thermal hydraulics modeling. Within the framework of the “Neutronics Phase” of the Benchmark the solutions of some selected test cases at the cell physics and lattice physics levels are presented. The SCALE 6.1 code package has been used for the neutronics modeling of the selected exercises. Sensitivity and Uncertainty analysis (S/U based on the generalized perturbation theory has been performed in order to assess the uncertainty of the computation of some selected reactor integral parameters due to the uncertainty in the basic nuclear data. As a general trend, it has been found that the main sources of uncertainty are the 238U (n, and the 239Pu nubar for the UOX- and the MOX-fuelled test cases, respectively. Moreover, the reference solutions for the test cases obtained using Monte Carlo methodologies together with a comparison between deterministic and stochastic solutions are presented.

Binary neutron star (BNS) mergers are the leading model to explain the phenomenology of short gamma-ray bursts (SGRBs), which are among the most luminous explosions in the universe. Recent observations of long-lasting X-ray afterglows of SGRBs challenge standard paradigms and indicate that in a large fraction of events a long-lived neutron star (NS) may be formed rather than a black hole. Understanding the mechanisms underlying these afterglows is necessary in order to address the open questions concerning the nature of SGRB central engines. However, recent theoretical progress has been hampered by the fact that the timescales of interest for the afterglow emission are inaccessible to numerical relativity simulations. Here we present a detailed model to bridge the gap between numerical simulations of the merger process and the relevant timescales for the afterglows, assuming that the merger results in a long-lived NS. This model is formulated in terms of a set of coupled differential equations that follow the...

Nuclear heating inside a MTR reactor has to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. This measurement is usually carried out by calorimetry. The innovative calorimetric system, CALMOS, has been studied and built in 2011 for the 70 MWth OSIRIS reactor operated by CEA. Thanks to a new type of calorimetric probe, associated to a specific displacement system, it provides measurements along the fissile height and above the core. This development required preliminary modelling and irradiation of mock-ups of the calorimetric probe in the ex-core area, where nuclear heating rate does not exceed 2 W.g{sup -1}. The calorimeter working modes, the different measurement procedures allowed with such a new probe, the main modeling and experimental results and expected advantages of this new technique have been already presented. However, these first in-core measurements were not performed beyond 6 W.g{sup -1}, due to an inside temperature limitation imposed by a safety authority requirement. In this paper, we present the first in-core simultaneous measurements of nuclear heating and conventional thermal neutronflux obtained by the CALMOS device at the 70 MW nominal reactor power. For the first time, this experimental system was operated in nominal in-core conditions, with nominal neutronflux up to 2.7 10{sup 14} n.cm{sup -2}.s{sup -1} and nuclear heating up to 12 W.g{sup -1}. A comprehensive measurement campaign carried out from 2013 to 2015 inside all accessible irradiation locations of the core, allowed to qualify definitively this new device, not only in terms of measurement ability but also in terms of reliability. After a brief reminder of the calorimetric cell configuration and displacement system specificities, first nuclear heating distributions at nominal power are presented and discussed. In order to reinforce the heating evaluation, a systematic comparison is made between results obtained by

Neutron spectrometry is a powerful method of investigating atomic nuclei and condensed matter. Such investigations provide necessary data for a very wide spectrum of scientific and technological applications from the fundamental problems of the structure of matter and nucleosynthesis in the Universe to atomic power technologies and the structure of condensed matter. The most frequently utilized is the time-of-flight (TOF) method for powerful pulsed neutron sources. However, in many cases, one can use more effective, simpler and cheaper methods. For example, for astrophysics and radioactive waste transmutation problems, it is sufficient to know an average resonance cross section or "resonance integrals" for capture and fission reactions for neutron spectra specific to neutronfluxes in stars or in the active zone of a transmutation reactor. In these cases, the slow-down neutron spectroscopy (SDNS) methods in lead and graphite moderators will be useful. Compared to the TOF method, the lead SDNS gives a 10/sup 3...

For approximately two decades, efforts have been sponsored by NASA's Marshall Space Flight Center to make possible high-speed, automated classification and quantification of constituent materials in various harsh environments. MSFC, along with the Air Force/Arnold Engineering Development Center, has led the work, developing and implementing systems that employ principles of emission and absorption spectroscopy to monitor molecular and atomic particulates in gas plasma of rocket engine flow fields. One such system identifies species and quantifies mass loss rates in H2/O2 rocket plumes. Other gases have been examined and the physics of their detection under numerous conditions were made a part of the knowledge base for the MSFC/USAF team. Additionally, efforts are being advanced to hardware encode components of the data analysis tools in order to address real-time operational requirements for health monitoring and management. NASA has a significant investment in these systems, warranting a spiral approach that meshes current tools and experience with technological advancements. This paper addresses current systems - the Optical Plume Anomaly Detector (OPAD) and the Engine Diagnostic Filtering System (EDIFIS) - and discusses what is considered a natural progression: a concept for migrating them towards detection of high energy particles, including neutrons and gamma rays. The proposal outlines system development to date, basic concepts for future advancements, and recommendations for accomplishing them.

For approximately two decades, efforts have been sponsored by NASA's Marshall Space Flight Center to make possible high-speed, automated classification and quantification of constituent materials in various harsh environments. MSFC, along with the Air Force/Arnold Engineering Development Center, has led the work, developing and implementing systems that employ principles of emission and absorption spectroscopy to monitor molecular and atomic particulates in gas plasma of rocket engine flow fields. One such system identifies species and quantifies mass loss rates in H2/O2 rocket plumes. Other gases have been examined and the physics of their detection under numerous conditions were made a part of the knowledge base for the MSFC/USAF team. Additionally, efforts are being advanced to hardware encode components of the data analysis tools in order to address real-time operational requirements for health monitoring and management. NASA has a significant investment in these systems, warranting a spiral approach that meshes current tools and experience with technological advancements. This paper addresses current systems - the Optical Plume Anomaly Detector (OPAD) and the Engine Diagnostic Filtering System (EDIFIS) - and discusses what is considered a natural progression: a concept for migrating them towards detection of high energy particles, including neutrons and gamma rays. The proposal outlines system development to date, basic concepts for future advancements, and recommendations for accomplishing them.

Binary neutron star (BNS) mergers are the leading model to explain the phenomenology of short gamma-ray bursts (SGRBs). Recent observations of long-lasting X-ray afterglows of SGRBs challenge standard paradigms and indicate that in a large fraction of events a long-lived neutron star (NS) may be formed rather than a black hole. Understanding the mechanisms underlying these afterglows is necessary in order to address the open questions concerning the nature of SGRB central engines. However, recent theoretical progress has been hampered by the fact that the timescales of interest for the afterglow emission are inaccessible to numerical relativity simulations. Here we present a detailed model to bridge the gap between numerical simulations of the merger process and the relevant timescales for the afterglows, assuming that the merger results in a long-lived NS. This model is formulated in terms of a set of coupled differential equations that follow the evolution of the post-merger system and predict its electromagnetic (EM) emission in a self-consistent way, starting from initial data that can be extracted from BNS merger simulations. The model presented here also allows us to search for suitable EM counterparts for multimessenger astronomy, which is expected to become reality within the next few years thanks to ground-based GW detectors such as advanced LIGO and Virgo. This paper discusses the formulation and implementation of the model. In a companion paper, we employ this model to predict the EM emission from ∼ {10}-2 to ∼ {10}7 {{s}} after a BNS merger and discuss the implications in the context of SGRBs and multimessenger astronomy.

A method for the solution of the inverse problem of radiative transfer is presented which utilizes the internal fluxes measured at different depths and in different directions with optical radiance microprobes in dense multiple scattering media. The method yields optical cross-sections and the ph...... Legendre coefficient of the unknown phase function can be recovered to the accuracy of 10%. Copyright (C) 1998 Elsevier Science Ltd. All rights reserved.......A method for the solution of the inverse problem of radiative transfer is presented which utilizes the internal fluxes measured at different depths and in different directions with optical radiance microprobes in dense multiple scattering media. The method yields optical cross......-sections and the phase function for the sample even when these parameters are depth dependent. The sensitivity analysis shows that the theoretical errors caused by the finite number of measurements as well as by the non-uniform directional sensitivity of the microprobes can be held on a low level; even the fourth...

A fail-safe, reliable, easy-to-use neutron shutter was designed, built, and put into operation at the Omega West Reactor, Los Alamos Scientific Laboratory. The neutron shutter will be used primarily to perform thermal neutron radiography, but is also available for a highly collimated source of thermal neutrons (neutronflux = 3.876 x 10/sup 6/ (neutrons)/(cm/sup 2/.s)). Neutron collimator sizes of either 10.16 by 10.16 cm or 10.16 by 30.48 cm are available.

The IPR-R1 is a reactor type TRIGA, Mark-I model, manufactured by the General Atomic Company and installed at Nuclear Technology Development Centre (CDTN), Brazilian Commission for Nuclear Energy (CNEN), in Belo Horizonte, Brazil. It is a light water moderated and cooled, graphite-reflected, open-pool type research reactor and operates at 100 kW. It presents low power, low pressure, for application in research, training and radioisotopes production. The fuel is an alloy of zirconium hydride and uranium enriched at 20% in {sup 235}U. The implementation of the PGNAA (Prompt Gamma Neutron Activation Analysis) using this research reactor will significantly increase in number of chemical elements analysed and the kind of matrices. A project is underway in order to implement this technique at CDTN. The objective of this study was to contribute in feasibility analysis of implementing this technique. For this purpose, MCNP is being used. Some variance reduction tools in the methodology, that has been already developed, was introduced for calculating of the neutronflux in the neutron extractor inclined. The objective was to reduce the code error and thereby increasing the reliability of the results. With the implementation of the variance reduction tools, the results of the thermal and epithermal neutronfluxes presented a significant improvement in both calculations. (author)

The CADIS and FW-CADIS hybrid Monte Carlo/deterministic techniques dramatically increase the efficiency of neutronics modeling, but their use in the accurate design analysis of very large and geometrically complex nuclear systems has been limited by the large number of processors and memory requirements for their preliminary deterministic calculations and final Monte Carlo calculation. Three mesh adaptivity algorithms were developed to reduce the memory requirements of CADIS and FW-CADIS without sacrificing their efficiency improvement. First, a macro-material approach enhances the fidelity of the deterministic models without changing the mesh. Second, a deterministic mesh refinement algorithm generates meshes that capture as much geometric detail as possible without exceeding a specified maximum number of mesh elements. Finally, a weight window coarsening algorithm de-couples the weight window mesh and energy bins from the mesh and energy group structure of the deterministic calculations in order to remove the memory constraint of the weight window map from the deterministic mesh resolution. The three algorithms were used to enhance an FW-CADIS calculation of the prompt dose rate throughout the ITER experimental facility. Using these algorithms resulted in a 23.3% increase in the number of mesh tally elements in which the dose rates were calculated in a 10-day Monte Carlo calculation and, additionally, increased the efficiency of the Monte Carlo simulation by a factor of at least 3.4. The three algorithms enabled this difficult calculation to be accurately solved using an FW-CADIS simulation on a regular computer cluster, obviating the need for a world-class super computer. (authors)

Nuclear reactors that rely upon the fission reaction have two modes of thermal energy deposition in the reactor system: neutron absorption and gamma absorption. The gamma rays are typically generated by neutron capture reactions or during the fission process which means the primary driver of energy production is of course the neutron interaction. In conventional reactor physics methods, the gamma heating component is ignored such that the gamma absorption is forced to occur at the gamma emission site. For experimental reactor systems like EBR-II and FFTF, the placement of structural pins and assemblies internal to the core leads to problems with power heating predictions because there is no fission power source internal to the assembly to dictate a spatial distribution of the power. As part of the EBR-II support work in the 1980s, the GAMSOR code was developed to assist analysts in calculating the gamma heating. The GAMSOR code is a modified version of DIF3D and actually functions within a sequence of DIF3D calculations. The gamma flux in a conventional fission reactor system does not perturb the neutronflux and thus the gamma flux calculation can be cast as a fixed source problem given a solution to the steady state neutronflux equation. This leads to a sequence of DIF3D calculations, called the GAMSOR sequence, which involves solving the neutronflux, then the gamma flux, and then combining the results to do a summary edit. In this manuscript, we go over the GAMSOR code and detail how it is put together and functions. We also discuss how to setup the GAMSOR sequence and input for each DIF3D calculation in the GAMSOR sequence.

One of the important neutron sources for Boron Neutron Capture Therapy (BNCT) is a nuclear reactor. It needs a high flux of epithermal neutrons. The optimum conditions of the neutron spectra for BNCT are provided by the International Atomic Energy Agency (IAEA). In this paper, Miniature Neutron Source Reactor (MNSR) as a neutron source for BNCT was investigated. For this purpose, we designed a Beam Shaping Assembly (BSA) for the reactor and the neutron transport from the core of the reactor t...

Highlights: • The in-vessel components of MFC system must withstand the extreme ITER environment. • To verify this, the thermal cycle test and the vibration tests were conducted. • Both tests were conducted under much severer conditions than ITER environment. • Soundness verification tests after the tests indicated that no problemswere found. • It is shown that the in-vessel component is sufficiently robust ITER environment. - Abstract: Via thermal cycling and vibration tests, this study aims to demonstrate that the in-vessel components of the microfission chamber (MFC) system can withstand the extreme International Thermonuclear Experimental Reactor (ITER) environment. In thermal cycle tests, the signal cable of the device was bent into a smaller radius and it was given more bends than those in its actual configuration within ITER. A faster rate of temperature change than that under the typical ITER baking scenario was then imposed on in-vessel components. For the vibration tests, strong 10 G vibrational accelerations with frequencies ranging from 30 Hz to 2000 Hz were imposed to the detector and the connector of the in-vessel components to simulate various types of electromagnetic events. Soundness verification tests of the in-vessel components conducted after thermal cycling and vibration testing indicated that problems related to the signal transmission cable functioning were not found. Thus, it was demonstrated that the in-vessel components of the MFC can withstand the extreme environment within ITER.

The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor safety. The general Monte Carlo radiation transport code MCNP, version 4B, has been used for these calculations.

This report summarizes the Workshop on Neutron Scattering Techniques for Studies in Catalysis, held at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) on September 16 and 17, 2010. The goal of the Workshop was to bring experts in heterogeneous catalysis and biocatalysis together with neutron scattering experimenters to identify ways to attack new problems, especially Grand Challenge problems in catalysis, using neutron scattering. The Workshop locale was motivated by the neutron capabilities at ORNL, including the High Flux Isotope Reactor (HFIR) and the new and developing instrumentation at the SNS. Approximately 90 researchers met for 1 1/2 days with oral presentations and breakout sessions. Oral presentations were divided into five topical sessions aimed at a discussion of Grand Challenge problems in catalysis, dynamics studies, structure characterization, biocatalysis, and computational methods. Eleven internationally known invited experts spoke in these sessions. The Workshop was intended both to educate catalyst experts about the methods and possibilities of neutron methods and to educate the neutron community about the methods and scientific challenges in catalysis. Above all, it was intended to inspire new research ideas among the attendees. All attendees were asked to participate in one or more of three breakout sessions to share ideas and propose new experiments that could be performed using the ORNL neutron facilities. The Workshop was expected to lead to proposals for beam time at either the HFIR or the SNS; therefore, it was expected that each breakout session would identify a few experiments or proof-of-principle experiments and a leader who would pursue a proposal after the Workshop. Also, a refereed review article will be submitted to a prominent journal to present research and ideas illustrating the benefits and possibilities of neutron methods for catalysis research.

The characterization of the energy distribution of a neutronflux is difficult in experiments with constrained geometry where techniques such as time of flight cannot be used to resolve the distribution. The measurement of neutronfluxes in reactors, which often present similar challenges, has been accomplished using radioactivation foils as an indirect probe. Spectral unfolding codes use statistical methods to adjust MCNP predictions of neutron energy distributions using quantified radioactive residuals produced in these foils. We have applied a modification of this established neutronflux characterization technique to experimentally characterize the neutronflux in the critical assemblies at the Nevada National Security Site (NNSS) and the spallation neutronflux at the Isotope Production Facility (IPF) at Los Alamos National Laboratory (LANL). Results of the unfolding procedure are presented and compared with a priori MCNP predictions, and the implications for measurements using the neutronfluxes at these facilities are discussed.

The fuel assembly or core with unstructured geometry is frequently used in the advanced reactor. To calculate the fuel assembly, the transmission probability method (TPM) is widely used. However, the rectangular or hexagonal meshes are mainly used in the TPM codes for the normal core structure. The triangle meshes are most useful for expressing the complicated unstructured geometry. Even though the finite element method and Monte-Carlo methodare well suited for solving the unstructured geometry problem, they are very time-consuming. Therefore, a TPM code based on the triangle meshes is developed here. This code was applied to the hybrid fuel geometry, and compared with the results of the MCNP code and other codes. The results of the comparison were consistent with each other. The TPM with triangle meshes can thus be applied to the two-dimensional arbitrary fuel assembly.

Recent work at Los Alamos Neutron Science Center (LANSCE), has shown that there are large gains in neutron beam intensity to be made by using coupled moderators at spallation neutron sources. Most of these gains result from broadening the pulse-width in time. However the accompanying longer exponential tail at large emission times can be a problem in that it introduces relatively large beam-related backgrounds at high resolutions. We have designed a device that can reshape the moderated neutron beam by cutting the time-tail so that a sharp time resolution can be re-established without a significant loss in intensity. In this work the basic principles behind the tail-cutter and some initial results of Monte Carlo simulations are described. Unwanted neutrons in the long time-tail are diffracted out of the transmitted neutron beam by a nested stack of aperiodic multi-layers, rocking at the same frequency as the source. Nested aperiodic multi-layers have recently been used at X-ray sources and as band-pass filters in quasi-Laue neutron experiments at reactor neutron sources. Optical devices that rock in synchronization with a pulsed neutron beam are relatively new but are already under construction at LANSCE. The tail-cutter described here is a novel concept that uses existing multi-layer technology in a new way for spallation neutrons. Coupled moderators in combination with beam shaping devices offer the means of increasing flux whilst maintaining a sharp time distribution. A prototype device is being constructed for the protein crystallography station at LANSCE. The protein crystallography station incorporates a water moderator that has been judiciously coupled in order to increase the flux over neutron energies that are important to structural biology (3-80meV). This development in moderator design is particularly important because protein crystallography is flux limited and because conventional ambient water and cold hydrogen moderators do not provide relatively

This paper involves the feasibility of boron neutron capture therapy (BNCT) for liver tumor with four sealed neutron generators as neutron source. Two generators are placed on each side of the liver. The high energy of these emitted neutrons should be reduced by designing a beam shaping assembly (BSA) to make them useable for BNCT. However, the neutronflux decreases as neutrons pass through different materials of BSA. Therefore, it is essential to find ways to increase the neutronflux. In this paper, the feasibility of using low enrichment uranium as a neutron multiplier is investigated to increase the number of neutrons emitted from D-T neutron generators. The neutron spectrum related to our system has a proper epithermal flux, and the fast and thermal neutronfluxes comply with the IAEA recommended values.

Boron-neutron capture therapy (BNCT) is a perspective treatment method for radiation resistant tumors. Unfortunately its development is strongly held back by a several physical and medical problems. Neutron sources for BNCT currently are limited to nuclear reactors and accelerators. For wide spread of BNCT investigations more compact and cheap neutron source would be much more preferable. In present paper an approach for compact D-D neutron generator creation based on a high current ECR ion source is suggested. Results on dense proton beams production are presented. A possibility of ion beams formation with current density up to 600 mA/cm2 is demonstrated. Estimations based on obtained experimental results show that neutron target bombarded by such deuteron beams would theoretically yield a neutronflux density up to 6·1010 cm-2/s. Thus, neutron generator based on a high-current deuteron ECR source with a powerful plasma heating by gyrotron radiation could fulfill the BNCT requirements significantly lower price, smaller size and ease of operation in comparison with existing reactors and accelerators.

Boron-neutron capture therapy (BNCT) is a perspective treatment method for radiation resistant tumors. Unfortunately its development is strongly held back by a several physical and medical problems. Neutron sources for BNCT currently are limited to nuclear reactors and accelerators. For wide spread of BNCT investigations more compact and cheap neutron source would be much more preferable. In present paper an approach for compact D–D neutron generator creation based on a high current ECR ion source is suggested. Results on dense proton beams production are presented. A possibility of ion beams formation with current density up to 600 mA/cm{sup 2} is demonstrated. Estimations based on obtained experimental results show that neutron target bombarded by such deuteron beams would theoretically yield a neutronflux density up to 6·10{sup 10} cm{sup −2}/s. Thus, neutron generator based on a high-current deuteron ECR source with a powerful plasma heating by gyrotron radiation could fulfill the BNCT requirements significantly lower price, smaller size and ease of operation in comparison with existing reactors and accelerators.

Nuclear reactors that rely upon the fission reaction have two modes of thermal energy deposition in the reactor system: neutron absorption and gamma absorption. The gamma rays are typically generated by neutron absorption reactions or during the fission process which means the primary driver of energy production is of course the neutron interaction. In conventional reactor physics methods, the gamma heating component is ignored such that the gamma absorption is forced to occur at the gamma emission site. For experimental reactor systems like EBR-II and FFTF, the placement of structural pins and assemblies internal to the core leads to problems with power heating predictions because there is no fission power source internal to the assembly to dictate a spatial distribution of the power. As part of the EBR-II support work in the 1980s, the GAMSOR code was developed to assist analysts in calculating the gamma heating. The GAMSOR code is a modified version of DIF3D and actually functions within a sequence of DIF3D calculations. The gamma flux in a conventional fission reactor system does not perturb the neutronflux and thus the gamma flux calculation can be cast as a fixed source problem given a solution to the steady state neutronflux equation. This leads to a sequence of DIF3D calculations, called the GAMSOR sequence, which involves solving the neutronflux, then the gamma flux, then combining the results to do a summary edit. In this manuscript, we go over the GAMSOR code and detail how it is put together and functions. We also discuss how to setup the GAMSOR sequence and input for each DIF3D calculation in the GAMSOR sequence. With the GAMSOR capability, users can take any valid steady state DIF3D calculation and compute the power distribution due to neutron and gamma heating. The MC2-3 code is the preferable companion code to use for generating neutron and gamma cross section data, but the GAMSOR code can accept cross section data from other sources. To further

PL equations are classical high order approximations to the transport equations which are based on the expansion of the angular dependence of the angular neutronflux and the nuclear cross sections in terms of spherical harmonics. A nodal collocation method is used to discretize the PL equations associated with a neutron source transport problem. The performance of the method is tested solving two 1D problems with analytical solution for the transport equation and a classical 2D problem. (authors)

X-ray diffraction using conventional laboratory equipment and/or synchrotron installations is the most important method for structure analyses. The purpose of this paper is to discuss special cases, for which, in addition to this indispensable part, neutrons are required to solve structural problems. Even though the huge intensity of modern synchrotron sources allows in principle the study of magnetic X-ray scattering the investigation of magnetic structures is still one of the most important applications of neutron diffraction. (author) 15 figs., 1 tab., 10 refs.

Selecting the best neutron source for boron neutron capture therapy (BNCT) requires optimizing neutron beam parameters. This involves solving many complex problems. Safety issues related to the use of nuclear reactor in hospital environments, as well as lower costs have led to interest in the development of accelerator-driven neutron sources. The BNCT research programs at the Nuclear Departments of Pisa and Genova Universities (DIMNP and DITEC) focus on studies of new concepts for accelerator-based DT neutron sources. Simple and compact accelerator designs using relatively low deuteron beam energy, ˜100 keV, have been developed which, in turn, can generate high neutron yields. New studies have been started for optimization of moderator materials for the 14.1 MeV DT neutrons. Our aim is to obtain an epithermal neutron beam for therapeutic application at the exit end, with minimal beam intensity losses, the specific goal is to achieve an epithermal neutronflux of at least of 1×109 n/cm2 s at the beam port, with low gamma and fast neutron dose contamination. According to the most recent neutron BNCT beam parameters some moderating and spectrum shifter materials and geometrical configurations have thus far been tested, and neutron and gamma beam data at beam port have been computed. A possible beam shaping assembly model has been designed. This research demonstrates that a DT neutron source could be successfully implemented for BNCT application, with performance surpassing the minimum requirements stated above, using DT neutron sources with yields in the range 1013-1014 n/s. The latest Monte Carlo simulation results of an accelerator based facility which relies on a rf-driven DT fusion neutron generator will be presented.

The international project MEGAPIE (MEGAwatt PIlot Experiment) at the Paul Scherrer Institute aims to build and operate the first 1 MW liquid lead-bismuth spallation target. This work is dedicated to the characterization of the neutronflux and the actinide incineration potential of the target. This mission has required the development of an innovating neutron detector (DNM) made of 8 micro fission chambers, installed inside the central rod of the MEGAPIE target. The combination of uranium chambers with chambers without deposit allows an efficient compensation of the gamma radiation background. The optimisation and development work on the MEGAPIE chambers have enabled us to measure the {sigma}{sub f} * {phi} product at each level of the DNM with an uncertainty of less than 3 per cent. We have inferred from these data the value of the epithermal neutronflux (E > 1 eV) at 37 cm away from the window: 3.4*10{sup 13} n.cm{sup -2}.s{sup -1}, and the values of the neutronflux at 50, 60 and 74 cm: 1.2*10{sup 13}, 7.9*10{sup 12} and 3.9*10{sup 12} n.cm{sup -2}.s{sup -1} respectively. All these values are notably less important than those obtained from MCNPX simulations. Thermocouples installed in DMN have enabled us to know the temperature distribution inside the target. For a beam intensity of 1.2 mA, the temperature ranges from 360 to 420 Celsius degrees in the low part of the central rod. The thermal inertia of the system composed of the central rod and DNM has been assessed for brutal changes of the beam intensity and is worth about 60 s. (A.C.)

Based on a three-dimensional modal geometry model for the WWER 440 reacotr, with nodes in the hexagonal z geometry, the equations for the interative calculation of the mean neutronflux density in a node and their variations due to stochastic control element vibration are shown. For modelling sources of noise, two different geometric and neutron-physics equations are used, according to the design of a control element as a spatial double pendulum with the absorber and fuel part. The neutronflux noise caused by vibration of the fuel parts is due to area sources. These are induced by material parameter variation due to control element displacement within the guide duct. The model of the `thermal black body` absorbing hollow cylinder is transferred to bodies of hexagonal crossection for the absorber part. Both sources of noise are described as disturbances for the partial neutron current densities averaged over the node surfaces in the two group diffusion approximation. The transfer of the noise signals is dealt with in the prompt response approximation. The `two group swelling nodes` are coupled to the `one group transmission nodes` on the basis of the modified one group diffusion approximation. The algorithms shown are the basis for development of a computer program for examining the transfer functions depending on location of neutronflux density variations with stochastic control element vibrations as the source of noise. (orig./HP) [Deutsch] Auf der Basis eines dreidimensionalen nodalen Geometriemodells fuer den WWER-440-Reaktor mit Nodes in Hexagonal-z-Geometrie werden die Beziehungen zur iterativen Berechnung der mittleren Neutronenflussdichte in einer Node sowie deren Schwankungen infolge stochastischer Regelelementschwingungen dargestellt. Fuer die Rauschquellenmodellierung werden entsprechend der Konstruktion eines Regelelements als raeumliches Doppelpendel mit Absorber- und Brennstoffteil zwei verschiedene geometrische und neutronenphysikalische Ansaetze

Modern neutron sources and science share a common origin in mid-20th-century scientific investigations concerned with the study of the fundamental interactions between elementary particles. Since the time of that common origin, neutron science and the study of elementary particles have evolved into quite disparate disciplines. The neutron became recognized as a powerful tool for studying condensed matter with modern neutron sources being primarily used (and justified) as tools for neutron scattering and materials science research. The study of elementary particles has, of course, led to the development of rather different tools and is now dominated by activities performed at extremely high energies. Notwithstanding this trend, the study of fundamental interactions using neutrons has continued and remains a vigorous activity at many contemporary neutron sources. This research, like neutron scattering research, has benefited enormously by the development of modern high-fluxneutron facilities. Future sources, particularly high-power spallation sources, offer exciting possibilities for continuing this research.

This paper presents experimental and computational techniques implemented for 4He gas scintillation detectors for induced fission neutron detection. Fission neutrons are produced when natural uranium samples are actively interrogated by 2.45 MeV deuterium-deuterium fusion reaction neutrons. Fission neutrons of energies greater than 2.45 MeV can be distinguished by their different scintillation pulse height spectra since 4He detectors retain incident fast neutron energy information. To enable the preferential detection of fast neutrons up to 10 MeV and suppress low-energy event counts, the detector photomultiplier gain is lowered and trigger threshold is increased. Pile-up and other unreliable events due to the interrogating neutronflux and background radiation are filtered out prior to the evaluation of pulse height spectra. With these problem-specific calibrations and data processing, the 4He detector's accuracy at discriminating fission neutrons up to 10 MeV is improved and verified with 252Cf spontaneous fission neutrons. Given the 4He detector's ability to differentiate fast neutron sources, this proof-of-concept active-interrogation measurement demonstrates the potential of special nuclear materials detection using a 4He fast neutron detection system.

The fractional point-neutron kinetics model for the dynamic behavior in a nuclear reactor is derived and analyzed in this paper. The fractional model retains the main dynamic characteristics of the neutron motion in which the relaxation time associated with a rapid variation in the neutronflux contains a fractional order, acting as exponent of the relaxation time, to obtain the best representation of a nuclear reactor dynamics. The physical interpretation of the fractional order is related with non-Fickian effects from the neutron diffusion equation point of view. The numerical approximation to the solution of the fractional neutron point kinetics model, which can be represented as a multi-term high-order linear fractional differential equation, is calculated by reducing the problem to a system of ordinary and fractional differential equations. The numerical stability of the fractional scheme is investigated in this work. Results for neutron dynamic behavior for both positive and negative reactivity and for different values of fractional order are shown and compared with the classic neutron point kinetic equations. Additionally, a related review with the neutron point kinetics equations is presented, which encompasses papers written in English about this research topic (as well as some books and technical reports) published since 1940 up to 2010.

We show data from a new type of detector that can be used to determine neutronflux, energy distribution, and direction of neutron motion for both fast and thermal neutrons. Many neutron detectors are plagued by large backgrounds from x-rays and gamma rays, and most current neutron detectors lack single-event energy sensitivity or any information on neutron directionality. Even the best detectors are limited by cosmic ray neutron backgrounds. All applications (neutron scattering and radiography, measurements of solar and cosmic ray neutronflux, measurements of neutron interaction cross sections, monitoring of neutrons at nuclear facilities, oil exploration, and searches for fissile weapons of mass destruction) will benefit from the improved neutron detection sensitivity and improved measurements of neutron properties made possible by this detector. The detector is free of backgrounds from x-rays, gamma rays, beta particles, relativistic singely charged particles and cosmic ray neutrons. It is sensitive to th...

Full Text Available Today facilities with collimated radiation field are widely used as reference in metrological support of devices for neutron radiation measurement. Neutron fields formed by radionuclide neutron sources. The aim of this research was to study characteristics of experimentally realized neutron fields geometries on АТ140 Neutron Calibration Facility using Monte Carlo method.For calibration, we put a device into neutron field with known flux density or ambient equivalent dose rate. We can form neutron beam from radionuclide fast-neutron source in different geometries. In containercollimator of АТ140 Neutron Calibration Facility we can install special inserts to gather fast-neutron geometry or thermal-neutron geometry. We need to consider neutron scattering from air and room’s walls. We can conduct measurements of neutron field characteristics in several points and get the other using Monte Carlo method.Thermal neutron collimator forms a beam from radionuclide source with a significant amount of neutrons with thermal energies. From found relationship between full neutronflux and distance to neutron source we see that inverse square law is violated. Scattered radiation contribution into total flux increases when we are moving away from neutron source and significantly influences neutron fields characteristics. While source is exposed in shadow-cone geometry neutron specter has pronounced thermal component from wall scattering.In this work, we examined main geometry types used to acquire reference neutron radiation using radionuclide sources. We developed Monte Carlo model for 238Pu-Be neutron source and АТ140 Neutron Calibration Facility’s container-collimator. We have shown the most significant neutron energy distribution factor to be scattered radiation from room’s walls. It leads to significant changes of neutron radiation specter at a distance from the source. When planning location, and installing the facility we should consider

Background: In the field of the treatment by Boron Neutron Capture Therapy (BNCT), an optimized neutron system was proposed. This study (simulation) was conducted to optimize the geometry and composition of neutron system and increase the epithermal neutronflux for the treatment of deep tumors is performed. Materials and Methods: A neutron system for BNCT was proposed. The system included 252Cf neutron source, neutron moderator/reflector arrangement, filter and concrete. To capture fast ...

AMTRAN, a one, two and three dimensional Sn neutron transport code with adaptive mesh refinement (AMR) has been parallelized with MPI over spatial domains and energy groups and with threads over angles. Block refined AMR is used with linear finite element representations for the fluxes, which are node centered. AMR requirements are determined by minimum mean free path calculations throughout the problem and can provide an order of magnitude or more reduction in zoning requirements for the same level of accuracy, compared to a uniformly zoned problem.

In the production of radioisotopes and neutron activation analysis, the fast neutron densities are very important to estimate the yields of the radioisotopes. In order to determine the fast neutronflux ratio, different foils are used to measure the thermal neutronflux and the fast neutronflux. In this paper 238U was used as only a monitor to measure the ratio of high energy neutron (>6 MeV). By measuring the

A low efficiency, 2-Dimensional Position Sensitive Neutron Detector based on delay line position encoding is developed. It is designed to handle beam flux of 106-107 n/cm2/s and for monitoring intensity profiles of neutron beams. The present detector can be mounted in transmission mode, as the hardware allows maximum neutron transmission in sensitive region. Position resolution of 1.2 mm in X and Y directions, is obtained. Online monitoring of beam images and intensity profile of various neutron scattering spectrometers at Dhruva are presented. It shows better dynamic range of intensity over commercial neutron camera and is also time effective over the traditionally used photographic method.

The McClellan Nuclear Radiation Center's (MNRC) staff in conjunction with a Cooperative Research and Development Agreement (CRDA) with the U.C. Santa Barbara facility has developed a system that can be used for aircraft inspection of jet engine blades. The problem was to develop an inspection system that can detect very low concentrations of hydrogen (i.e., greater than 100 ppm) in metal matricies. Specifically in Titanium alloy jet engine blades. Entrapment and precipitation of hydrogen in metals is an undesirable phenomenon which occurs in many alloys of steel and titanium. In general, metals suffer a loss of mechanical properties after long exposures to hydrogen, especially at high temperatures and pressures, thereby becoming embrittled. Neutron radiography has been used as a nondestructive testing technique for many years. Neutrons, because of their unique interactions with materials, are especially useful in the detection of hydrogen. They have an extremely high interaction cross section for low atomic number nuclei (i.e., hydrogen). Thus hydrogen in a metal matrix can be visualized using neutrons. Traditional radiography is sensitive to the total attenuation integrated over the path of radiation through the material. Increased sensitivity and quantitative cross section resolution can be obtained using three-dimensional volumetric imaging techniques such as tomography. The solution used to solve the problem was to develop a neutron tomography system. The neutron source is the McClellan Nuclear Radiation Center's 1 MW TRIGA reactor. This paper describes the hardware used in the system as well as some of the preliminary results.

The PCS (Protein Crystallography Station) at Los Alamos Neutron Science Center (LANSCE) is a unique facility in the USA that is designed and optimized for detecting and collecting neutron diffraction data from macromolecular crystals. PCS utilizes the 20 Hz spallation neutron source at LANSCE to enable time-of-flight measurements using 0.6-7.0 {angstrom} neutrons. This increases the neutronflux on the sample by using a wavelength range that is optimal for studying macromolecular crystal structures. The diagram below show a schematic of PCS and photos of the detector and instrument cave.

Systematic measurements of the concentrations of cosmogenic 41Ca (half-life = 1.04 × 10 5 yr) in the Apollo 15 long core 15001-15006 were performed by accelerator mass spectroscopy. Earlier measurements of cosmogenic 10Be, 14C, 26Al, 36Cl, and 53Mn in the same core have provided confirmation and improvement of theoretical models for predicting production profiles of nuclides by cosmic ray induced spallation in the Moon and large meteorites. Unlike these nuclides, 41Ca in the lunar surface is produced mainly by thermal neutron capture reactions on 40Ca. The maximum productions of 41Ca, about 1 dpm/g Ca, was observed at a depth in the Moon of about 150 g/cm 2. For depths below about 300 g/cm 2, 41Ca production falls off exponentially with an e-folding length of 175 g/cm 2. Neutron production in the Moon was modeled with the Los Alamos High Energy Transport Code System, and yields of nuclei produced by low-energy thermal and epithermal neutrons were calculated with the Monte Carlo N-Particle code. The new theoretical calculations using these codes are in good agreement with our measured 41Ca concentrations as well as with 60Co and direct neutron fluence measurements in the Moon.

Based on a new second-order neutron transport equation, self-adjoint angular flux (SAAF) equation, the spherical harmonics (PN) method for neutron transport equation on unstructured-meshes is derived. The spherical harmonics function is used to expand the angular flux. A set of differential equations about the spatial variable, which are coupled with each other, can be obtained. They are solved iteratively by using the finite element method on unstructured-meshes. A two-dimension transport calculation program is coded according to the model. The numerical results of some benchmark problems demonstrate that this method can give high precision results and avoid the ray effect very well.

This report is a supplement to the original report, published in 1968, as revised. The Benchmark Problem Book is intended to serve as a source book of solutions to mathematically well-defined problems for which either analytical or very accurate approximate solutions are known. This supplement contains problems in eight new areas: two-dimensional (R-z) reactor model; multidimensional (Hex-z) HTGR model; PWR thermal hydraulics--flow between two channels with different heat fluxes; multidimensional (x-y-z) LWR model; neutron transport in a cylindrical ''black'' rod; neutron transport in a BWR rod bundle; multidimensional (x-y-z) BWR model; and neutronic depletion benchmark problems. This supplement contains only the additional pages and those requiring modification. (RWR)

Monte Carlo simulations were carried out to design a prompt {\\gamma}-ray neutron activation analysis (PGNAA) thermal neutron output setup using MCNP5 computer code. In these simulations the moderator materials, reflective materials and structure of the PGNAA 252Cf neutrons of thermal neutron output setup were optimized. Results of the calcuations revealed that the thin layer paraffin and the thick layer of heavy water moderated effect is best for 252Cf neutrons spectrum. The new design compared with the conventional neutron source design, the thermal neutronflux and rate were increased by 3.02 times and 3.27 times. Results indicate that the use of this design should increase the neutronflux of prompt gamma-ray neutron activation analysis significantly.

Eddy correlation techniques to determine the turbulent fluxes of heat, moisture and momentum in the near-surface atmospheric layer rely on the Monin-Obukhov similarity theory, which requires stationarity and horizontal homogeneity. Experiments at specially selected sites over land and particularly over sea are used to develop this concept. Recent experiments, deliberately conducted in non-ideal conditions, show an underestimation of turbulent fluxes. Results from the field experiments FIFE, KUREX, TARTEX and SADE, point to a relationship between the underestimation of turbulent fluxes and terrain inhomogeneity. In order to systematically correct for this effect a scheme is suggested which uses fetch lengths of different types of surface in the sites surrounding the environment. In addition, horizontal differences in atmospheric stability above different surfaces are included in the correction scheme. This scheme might be useful for the design of validation experiments in non-homogeneous terrain.

With the purpose of updating knowledge in the fields of cold neutron sources, the work of this thesis has been run according to the 3 following axes. First, the gathering of specific information forming the materials of this work. This set of knowledge covers the following fields: cold neutron, cross-sections for the different cold moderators, flux slowing down, different measurements of the cold flux and finally, issues in the thermal analysis of the problem. Secondly, the study and development of suitable computation tools. After an analysis of the problem, several tools have been planed, implemented and tested in the 3-dimensional radiation transport code Tripoli-4. In particular, a module of uncoupling, integrated in the official version of Tripoli-4, can perform Monte-Carlo parametric studies with a spare factor of Cpu time fetching 50 times. A module of coupling, simulating neutron guides, has also been developed and implemented in the Monte-Carlo code McStas. Thirdly, achieving a complete study for the validation of the installed calculation chain. These studies focus on 3 cold sources currently functioning: SP1 from Orphee reactor and 2 other sources (SFH and SFV) from the HFR at the Laue Langevin Institute. These studies give examples of problems and methods for the design of future cold sources.

Full Text Available The neutron diffusion equation is often used to perform core-level neutronic calculations. It consists of a set of second-order partial differential equations over the spatial coordinates that are, both in the academia and in the industry, usually solved by discretizing the neutron leakage term using a structured grid. This work introduces the alternatives that unstructured grids can provide to aid the engineers to solve the neutron diffusion problem and gives a brief overview of the variety of possibilities they offer. It is by understanding the basic mathematics that lie beneath the equations that model real physical systems; better technical decisions can be made. It is in this spirit that this paper is written, giving a first introduction to the basic concepts which can be incorporated into core-level neutronflux computations. A simple two-dimensional homogeneous circular reactor is solved using a coarse unstructured grid in order to illustrate some basic differences between the finite volumes and the finite elements method. Also, the classic 2D IAEA PWR benchmark problem is solved for eighty combinations of symmetries, meshing algorithms, basic geometric entities, discretization schemes, and characteristic grid lengths, giving even more insight into the peculiarities that arise when solving the neutron diffusion equation using unstructured grids.

For nuclear reactor analysis such as the neutron eigenvalue calculations, the time consuming Monte Carlo (MC) simulations can be accelerated by using graphics processing units (GPUs). However, traditional MC methods are often history-based, and their performance on GPUs is affected significantly by the thread divergence problem. In this paper we describe the development of a newly designed event-based vectorized MC algorithm for solving the neutron eigenvalue problem. The code was implemented using NVIDIA's Compute Unified Device Architecture (CUDA), and tested on a NVIDIA Tesla M2090 GPU card. We found that although the vectorized MC algorithm greatly reduces the occurrence of thread divergence thus enhancing the warp execution efficiency, the overall simulation speed is roughly ten times slower than the history-based MC code on GPUs. Profiling results suggest that the slow speed is probably due to the memory access latency caused by the large amount of global memory transactions. Possible solutions to improve the code efficiency are discussed.

Neutrons are moderated or reduced in energy by scattering off of nuclei. When cosmic neutrons with high kinetic energy enter earth’s atmosphere...neutronflux. The simulation volume was modeled as a sphere centered at the origin with a radius of 100 cm. The shielding material was modeled as a

A technical assessment of the critical issues and problem areas for high heat flux materials and components (HHFMC) in magnetic fusion devices shows these problems to be of critical importance for the successful operation of near-term fusion experiments and for the feasibility and attractiveness of long-term fusion reactors. A number of subgroups were formed to assess the critical HHFMC issues along the following major lines: (1) source conditions, (2) systems integration, (3) materials and processes, (4) thermal hydraulics, (5) thermomechanical response, (6) electromagnetic response, (7) instrumentation and control, and (8) test facilities. The details of the technical assessment are presented in eight chapters. The primary technical issues and needs for each area are highlighted.

In this work it is presented the use of CR-39 detector to determine the thermal neutronflux distribution in a thermal facility of RP-10 and other critical places of interest. The method consist of radiating the CR-39 detectors (TASTRAK, 500 mm of thickness and area 1 cm{sup 2}) within an acrylic container placed to different depths inside thermal column. The irradiation time went of 9 h at power 10 MW. The chemical etching was carried out using a solution of NaOH 6.25 N to 71 deg C for 6 h. The number of damages (tracks) was counted visually with an optical microscope (Carl Zeiss to 160 increases). Finally, the obtained results were compared with those obtained ones by mean of the activation technique, showing discrepancies of 5% in all length of thermal facility. (author)

The National Institute of Nuclear Research (ININ) has in its installations with a nuclear subcritical reactor which was designed and constructed with the main purpose to be used in the nuclear sciences education in the Physics areas and Reactors engineering. Within the nuclear experiments that can be realized in this reactor are very interesting those about determinations of neutron and gamma fluxes spectra, since starting from these some interesting nuclear parameters can be obtained. In order to carry out this type of experiments different radioactive sources are used which exceed the permissible doses by far to human beings. Therefore it is necessary the remote handling as of the source as of detectors used in different experiments. In this work it is presented the design of an electronic system which allows the different positions inside of the tank of subcritical reactor at ININ over the radial and axial axes in manual or automatic ways. (Author)

In this paper, a sort of non-parameter model is constructed with the unit step response of the nuclear reactor neutron kinetics system. Furthermore, a sort of constant neutronflux density control law is presented using the double values dynamic matrix control principle. In contrast to the other control methods based on the accurate model, the method presented in this paper has good tracking performance and robustness. It can work despite the existence of un-measurable disturbances. The simulation experiment testifies the correctness and effectiveness of the method.%利用核反应堆中子动力学系统的单位阶跃响应数据，获得了该系统的非参数模型，且提出了一种中子通量密度恒值问题的双值动态矩阵控制新方法。与基于精确模型的控制方法相比，此方法不必苛求模型的具体形式，且实时控制的计算量小、跟踪调节性能好、鲁棒性强、能消除不可测干扰。仿真结果验证了这种控制律的有效性和优越性。

Neutron dosimetry measurements have been conducted for various positions of the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) in order to measure the neutronflux and energy spectra. Neutron dosimetry results and radiation damage calculations are presented for positions V10, V14, and V15.

For the first three decades following World War II, the US, which pioneered the field of neutron scattering research, enjoyed uncontested leadership in the field. By the mid-1970's, other countries, most notably through the West European consortium at Institut Laue-Langevin (ILL) in Grenoble, France, had begun funding neutron scattering on a scale unmatched in this country. By the early 1980's, observers charged with defining US scientific priorities began to stress the need for upgrading and expansion of US research reactor facilities. The conceptual design of the ANS facility is now well under way, and line-item funding for more advanced design is being sought for FY 1992. This should lead to a construction request in FY 1994 and start-up in FY 1999, assuming an optimal funding profile. While it may be too early to finalize designs for instruments whose construction is nearly a decade removed, it is imperative that we begin to develop the necessary concepts to ensure state-of-the-art instrumentation for the ANS. It is in this context that this Instrumentation Workshop was planned. The workshop touched upon many ideas that must be considered for the ANS, and as anticipated, several of the discussions and findings were relevant to the planning of the HFBR Upgrade. In addition, this report recognizes numerous opportunities for further breakthroughs on neutron instrumentation in areas such as improved detection schemes (including better tailored scintillation materials and image plates, and increased speed in both detection and data handling), in-beam monitors, transmission white beam polarizers, multilayers and supermirrors, and more. Each individual report has been cataloged separately.

The study of the dynamic behavior of next generation nuclear reactors is a fundamental aspect for safety and reliability assessments. Despite the growing performances of modern computers, the full solution of the neutron Boltzmann equation in the time domain is still an impracticable task, thus several approximate dynamic models have been proposed for the simulation of nuclear reactor transients; the quasi-static method represents the standard tool currently adopted for the space-time solution of neutron transport problems. All the practical applications of this method that have been proposed contain a major limit, consisting in the use of isotropic quantities, such as scalar fluxes and isotropic external neutron sources, being the only data structures available in most deterministic transport codes. The loss of the angular information produces both inaccuracies in the solution of the kinetic model and the inconsistency of the quasi-static method itself. The present paper is devoted to the implementation of a consistent quasi-static method. The computational platform developed by CEA in Cadarache has been used for the creation of a kinetic package to be coupled with the existing SNATCH solver, a discrete-ordinate multi-dimensional neutron transport solver, employed for the solution of the steady-state Boltzmann equation. The work aims at highlighting the effects of the angular treatment of the neutronflux on the transient analysis, comparing the results with those produced by the previous implementations of the quasi-static method. (author)

Using the Rossi X-ray Timing Explorer we have measured lags of the 9 to 33 keV photons relative to the 2 to 9 keV photons in the timing noise between 0.01 and 100 Hz in the accreting neutron stars 4U 0614+091 and 4U 1705-44. We performed similar measurements on the accreting black hole candidates Cyg X-1 and GX 339-4 as a comparison. During the observations these sources were all in low (hard) states. We find phase lags of between 0.03 and 0.2 radians in all these sources, with a variation in...

A central problem in nuclear reactor analysis is calculating solutions of steady-state k-eigenvalue problems with thermal hydraulic feedback. In this paper we propose and utilize a model problem that permits the theoretical analysis of iterative schemes for solving such problems. To begin, we discuss a model problem (with nonlinear cross section feedback) and its justification. We proceed with a Fourier analysis for source iteration schemes applied to the model problem. Then we analyze commonly-used iteration schemes involving non-linear diffusion acceleration and feedback. For each scheme we show (1) that they are conditionally stable, (2) the conditions that lead to instability, and (3) that traditional relaxation approaches can improve stability. Lastly, we propose a new iteration scheme that theory predicts is an improvement upon the existing methods.

Coded aperture techniques have been applied to neutron radiography to address limitations in neutronflux and resolution of neutron detectors in a system labeled coded source imaging (CSI). By coding the neutron source, a magnified imaging system is designed with small spot size aperture holes (10 and 100 m) for improved resolution beyond the detector limits and with many holes in the aperture (50% open) to account for flux losses due to the small pinhole size. An introduction to neutron radiography and coded aperture imaging is presented. A system design is developed for a CSI system with a development of equations for limitations on the system based on the coded image requirements and the neutron source characteristics of size and divergence. Simulation has been applied to the design using McStas to provide qualitative measures of performance with simulations of pinhole array objects followed by a quantitative measure through simulation of a tilted edge and calculation of the modulation transfer function (MTF) from the line spread function. MTF results for both 100um and 10um aperture hole diameters show resolutions matching the hole diameters.

Coded aperture techniques have been applied to neutron radiography to address limitations in neutronflux and resolution of neutron detectors in a system labeled coded source imaging (CSI). By coding the neutron source, a magnified imaging system is designed with small spot size aperture holes (10 and 100μm) for improved resolution beyond the detector limits and with many holes in the aperture (50% open) to account for flux losses due to the small pinhole size. An introduction to neutron radiography and coded aperture imaging is presented. A system design is developed for a CSI system with a development of equations for limitations on the system based on the coded image requirements and the neutron source characteristics of size and divergence. Simulation has been applied to the design using McStas to provide qualitative measures of performance with simulations of pinhole array objects followed by a quantitative measure through simulation of a tilted edge and calculation of the modulation transfer function (MTF) from the line spread function. MTF results for both 100μm and 10μm aperture hole diameters show resolutions matching the hole diameters.

The real-time and accurate measurement of reactor neutronflux is directly related to safe operation of the reactor.Neutronflux measurement system with fission chamber based on the analog technology had some shortage, a fission chamber of digital output signal processing system was designed.The combination of two kinds of fission chamber working mode which is pulse and Campbell were applied in this system.Using high-speed ADC digitalized signal from the fission chamber, FPGA could be used in a variety of digital signal pro-cessing algorithms.And using MATLAB simulated output pulse signal of the fission chamber, which imported arbitrary waveform signal generator as neutron source, to be used in a preliminary validation of the algorithm. Test results showed that the digital processing system had good linearity and good performance.%实时准确地测量反应堆的中子通量变化状况，对于确保反应堆的安全运行有着重要意义。基于模拟测量系统存在的一些不足，设计一种数字化的宽量程裂变室输出信号处理系统。该数字化信号处理系统将裂变室的脉冲和坎贝尔两种工作模式相结合，利用高速ADC对裂变室输出脉冲信号数字化，在FPGA中运用数字信号处理算法实现两种工作模式的信号处理，并仿真了裂变室的输出脉冲信号导入任意波形信号发生器模拟中子信号源，用于对处理算法的初步验证。测试结果表明数字化处理系统线性度和性能良好。

The intensity and spectrum of solar neutrons in the vicinity of the earth are calculated on the assumption that the low-energy protons recently detected in balloon and satellite flights are products of solar neutron decay. The solar-neutronflux thus obtained exceeds the global average cosmic-ray neutron leakage above 10 Mev, indicating that it may be an important source of both the inner and outer radiation belts. Neutron measurements in the atmosphere are reviewed and several features of the data are found to be consistent with the estimated solar neutron spectrum.

The huge neutronfluxes offer the possibility to use research reactors to produce isotopes of interest, which can be investigated afterwards. An example is the half-lives of long-lived isotopes like 129I. A direct usage of reactor neutrons in the astrophysical energy regime is only possible, if the corresponding ions are not at rest in the laboratory frame. The combination of an ion storage ring with a reactor and a neutron guide could open the path to direct measurements of neutron-induced cross sections on short-lived radioactive isotopes in the astrophysically interesting energy regime.

Neutrons in stable nuclei can exist forever; a free neutron lasts for about 15 minutes on average before it beta decays to a proton, an electron, and an antineutrino. Precision measurements of the neutron lifetime test the validity of weak interaction theory and provide input into the theory of the evolution of light elements in the early universe. There are two predominant ways of measuring the neutron lifetime: the bottle method and the beam method. The bottle method measures decays of ultracold neutrons that are stored in a bottle. The beam method measures decay protons in a beam of cold neutrons of known flux. An improved beam experiment is being prepared at the National Institute of Science and Technology (Gaithersburg, MD) with the goal of reducing statistical and systematic uncertainties to the level of 1 s. The purpose of my studies was to develop computer simulations of neutron transport to determine the beam collimation and study the neutron distribution's effect on systematic effects for the experiment, such as the solid angle of the neutronflux monitor. The motivation for the experiment and the results of this work will be presented. This work was supported, in part, by a Grant to Gettysburg College from the Howard Hughes Medical Institute through the Precollege and Undergraduate Science Education Program.

In this work the thermal neutrons diffusion is studied with interfacial effects in a fuel-moderator arrangement that consist of an infinite series of plane fuel plates and of moderator willing so that each plate of multiplicative material has in each end a moderator plate. The developed pattern is an unidimensional model for the thermal group obtained of equation of volumetric diffusion average. One analysis of parametric sensibility was realized to find the correction constants for the diffusion coefficient, the absorption term and the new transfer or current term in the fuel-moderator interface. The obtained results are compared against the classic theory, being obtained a good agreement among both theories. (Author)

A description of D-T neutron generator (NG) is presented. This machine can be used for fast neutron activation analysis applied to determine some selected elements, especially light elements, in different materials. Procedure of neutronflux determination and efficiency calculation is described. Examples of testing some Egyptian natural cosmetics are given.

The gas-discharge detector is designed for the neutron lifetime spectrometer. The detector is intended for ultracold neutronflux monitoring in measurement cycles at the specrtometer (ILL, Grenoble, France). The detector has been successively tested with a Pu-Be neutron source under laboratory conditions and as a part of the spectrometer.

Burnable poison rods, made of B{sub 4}C- Al{sub 2} O{sub 3} pellets with 5.01 mg/cm{sup 3} {sup 10} B concentration, have been manufactured for a set of experiments in the IPEN/MB-01 zero-power reactor. Several core parameters which are affected by the burnable poisons rods have been measured. The principal results, for the situation in which the burnable poison rods are located near the absorber rods of a control rod, are they cause a 29% rod worth shadowing, a reduction of 39% in the local void coefficient of reactivity, a reduction of 4.8% in the isothermal temperature coefficient of reactivity, and a reduction of 9% in the thermal neutronflux in the region where the burnable poison rods are located. These experimental results will be used for the validation of burnable poison calculation methods in the CTMSP. (author)

Full Text Available The presence of fast neutron spectra in new reactors is expected to induce a strong impact on the contained materials, including structural materials, nuclear fuels, neutron reflecting materials, and tritium breeding materials. Therefore, introduction of these reactors into operation will require extensive testing of their components, which must be performed under neutronic conditions representative of those expected to prevail inside the reactor cores when in operation. Due to limited availability of fast reactors, testing of future reactor materials will mostly take place in water cooled material test reactors (MTRs by tailoring the neutron spectrum via neutron screens. The latter rely on the utilization of materials capable of absorbing neutrons at specific energy. A large but fragmented experience is available on that topic. In this work a comprehensive compilation of the existing neutron screen technology is attempted, focusing on neutron screens developed in order to locally enhance the fast over thermal neutronflux ratio in a reactor core.

and MEGAPIE has been correctly reproduced. To achieve a good accuracy in the calculation of the neutronic performance of an ADS system, an accurate definition of the geometrical model taking into account the influence of structural materials is of primary importance. The results depend also on the beam profile used in the simulations, at least for the flux calculations close to the target interaction point. Radioactive nuclides produced in liquid metal targets are transported into hot cells, pumps or close to electronics with radiation sensitive components. Besides the considerable amount of decay {gamma} activity in the irradiated liquid metal, a significant amount of the Delayed Neutron (DN) precursor activity accumulates in the target fluid. The transit time of a liquid metal target being as short as a few seconds, DNs may contribute significantly to the activation and dose rates. The importance of the DN issues in liquid metal targets is confirmed. Another problem is the gas production and release in an ADS target, the proton beam generating a large amount of gas by spallation reactions. A large amount of Po isotopes, volatile at relatively high temperatures, are produced in the LBE. The gas production was measured by {gamma} spectroscopy. The release rates of noble gases in MEGAPIE are at the % level after 1-2 days of operation, while the release becomes almost complete weeks after the beginning of operation. Pressure increase in the cover gas could be reproduced with calculations within a factor of 2. The effect of the impurities in the radionuclide inventory of the LBE, using the actual chemical composition of the LBE used in MEGAPIE, is minimal.

Full Text Available Background: In the field of the treatment by Boron Neutron Capture Therapy (BNCT, an optimized neutron system was proposed. This study (simulation was conducted to optimize the geometry and composition of neutron system and increase the epithermal neutronflux for the treatment of deep tumors is performed. Materials and Methods: A neutron system for BNCT was proposed. The system included 252Cf neutron source, neutron moderator/reflector arrangement, filter and concrete. To capture fast neutrons, different neutron filters Fe, Pb, Ni and PbF2 with various thicknesses were simulated and studied. Li (with 1 mm thick was used for filtering of thermal neutrons. Bi with thickness of 1 cm was used to minimize the intensity of gamma rays. Monte Carlo simulation code MCNPX 2.4.0 was used for design of the neutron system and calculation of the neutron components at the output port of the system. Results: For different thicknesses of the filters, the fast neutronflux, the epithermal and thermal flux were calculated at the output port of the system. The spatial distribution of the fast neutronflux, the epithermal flux and gamma flux in human head phantom with the presence of 40 ppm of 10B were obtained. The present calculations showed that Pb filter (about 1 cm at the output port is suitable for fast neutron capture. The thickness of Li filter was determined due to its high absorption cross-section in thermal region. Bi was used as a gamma filter by the reason of it is good for shielding gamma rays, while having high transmission epithermal neutrons. Conclusion: The epithermal neutronflux has enhanced about 38 percent at the output port of the present system, compared with recent system proposed by Ghassoun et al. At 2 cm depth inside the head phantom, the neutronflux reaches a maximum value about . At this depth, the ratio of the thermal neutronflux to the epithermal flux is about three times, that suggests such a neutron system to treat tumors in the

A review is presented of those properties of neutron stars upon which there is general agreement and of those areas which currently remain in doubt. Developments in theoretical physics of neutron star interiors are summarized with particular attention devoted to hyperon interactions and the structure of interior layers. Determination of energy states and the composition of matter is described for successive layers, beginning with the surface and proceeding through the central region into the core. Problems encountered in determining the behavior of matter in the ultra-high density regime are discussed, and the effects of the magnetic field of a neutron star are evaluated along with the behavior of atomic structures in the field. The evolution of a neutron star is outlined with discussion centering on carbon detonation, cooling, vibrational damping, rotation, and pulsar glitches. The role of neutron stars in cosmic-ray propagation is considered.

Emplacement of contaminated reactor components involves disposal in lined and unlined auger holes in soil above the water table. The radionuclide inventory of disposed components was calculated. Information on the composition and weight of the components, as well as reasonable assumptions for the neutronflux fueling use, the time of neutron exposure, and radioactive decay after discharge, were employed in the inventory calculation. Near-field release rates of /sup 152/Eu, /sup 154/Eu, and /sup 155/Eu from control plates and cylinders were calculated for 50 years after emplacement. Release rates of the europium isotopes were uncertain. Two release-rate-limiting models were considered and a range of reasonable values were assumed for the time-to-failure of the auger-hole linear and aluminum cladding and europium solubility in SWSA-6 groundwater. The bounding europium radionuclide near-field release rates peaked at about 1.3 Ci/year total for /sup 152,154,155/Eu in 1987 for the lower bound, and at about 420 Ci/year in 1992 for the upper bound. The near-field release rates of /sup 55/Fe, /sup 59/Ni, /sup 60/Co, and /sup 63/Ni from stainless steel and cobalt alloy components, as well as of /sup 10/Be, /sup 41/Ca, and /sup 55/Fe from beryllium reflectors, were calculated for the next 100 years, assuming bulk waste corrosion was the release-rate-limiting step. Under the most conservative assumptions for the reflectors, the current (1986) total radionuclide release rate was calculated to be about 1.2 x 10/sup -4/ Ci/year, decreasing by 1992 to a steady release of about 1.5 x 10/sup -5/ Ci/year due primarily to /sup 41/Ca. 50 refs., 13 figs., 8 tabs.

In some applications of perturbation theory, it is necessary know the adjoint neutronflux, which is obtained by the solution of adjoint neutron diffusion equation. However, the multigroup constants used for this are weighted in only the direct neutronflux, from the solution of direct P1 equations. In this work, this procedure is questioned and the adjoint P1 equations are derived by the neutron transport equation, the reversion operators rules and analogies between direct and adjoint parameters. (author)

The main goal of the thesis about studies of cold neutrons sources (CNS) in research reactors was to create a complete set of tools to design efficiently CNS. The work raises the problem to run accurate simulations of experimental devices inside reactor reflector valid for parametric studies. On one hand, deterministic codes have reasonable computation times but introduce problems for geometrical description. On the other hand, Monte Carlo codes give the possibility to compute on precise geometry, but need computation times so important that parametric studies are impossible. To decrease this computation time, several developments were made in the Monte Carlo code TRIPOLI-4.4. An uncoupling technique is used to isolate a study zone in the complete reactor geometry. By recording boundary conditions (incoming flux), further simulations can be launched for parametric studies with a computation time reduced by a factor 60 (case of the cold neutron source of the Orphee reactor). The short response time allows to lead parametric studies using Monte Carlo code. Moreover, using biasing methods, the flux can be recorded on the surface of neutrons guides entries (low solid angle) with a further gain of running time. Finally, the implementation of a coupling module between TRIPOLI- 4.4 and the Monte Carlo code McStas for research in condensed matter field gives the possibility to obtain fluxes after transmission through neutrons guides, thus to have the neutronflux received by samples studied by scientists of condensed matter. This set of developments, involving TRIPOLI-4.4 and McStas, represent a complete computation scheme for research reactors: from nuclear core, where neutrons are created, to the exit of neutrons guides, on samples of matter. This complete calculation scheme is tested against ILL4 measurements of flux in cold neutron guides. (authors)

A discrete ordinates method for a threedimensional first-order neutron transport equation based on unstructured-meshes that avoids the singularity of the second-order neutron transport equation in void regions was derived.The finite element variation equation was obtained using the least-squares method.A three-dimensional transport calculation code was developed.Both the triangular-z and the tetrahedron elements were included.The numerical results of some benchmark problems demonstrated that this method can solve neutron transport problems in unstructuredmeshes very well.For most problems,the error of the eigenvalue and the angular flux is less than 0.3% and 3.0% respectively.

process, effectiveness of them are studied for the enhancement of the available system. Final conclusion from the simulation process, indicates that the heavy water with the thickness of 50 to 60 cm. is the best moderator for gaining the better thermal neutronsflux for enhancement of P.N.D. in the T.R.R. Powder Neutron Diffractometer y (P.N.D.) is relatively good and practical way for identification of the 3 dimensional construction of materials. In order to exploit the capabilities of this method, in one of the neutron beam of the Tehran Research Reactor (T.R.R.), a collimator embedded inside the concrete wall, direct the neutrons produced in the core reactor towards a monochromator e. Neutrons having been monochromated by 2 nd collimator are then directed towards the sample. Then the pattern of diffracted neutrons from the sample are studied. In order to make the best out of it, neutrons coming to sit on the sample must be of the thermal type. That means the number/amount of thermal neutronsflux in compar...

作为获国际认可的第四代核电站反应堆堆型之一的熔盐堆(Molten salt reactor,MSR),具有固有安全性高、经济性好、核资源可持续发展以及易于防止核扩散等优点.针对石墨-熔盐零功率堆的几何参数,利用蒙特卡罗计算程序MCNP5建立了物理计算模型,计算临界情况下堆芯径向、轴向中子通量及增殖区厚度与Th-U转换比(Conversion ratio,CR)的关系.结果表明,(1)石墨-熔盐零功率堆堆芯中子通量密度分布较为平坦；(2)石墨-熔盐零功率堆反射层厚度和增殖区厚度在一定范围内,CR随反射层厚度或增殖区厚度的增加而增加,当超出该范围,CR不再随反射层厚度或增殖区厚度的增加而明显增加.%The molten salt reactor (MSR) is the only one liquid-fuel reactor in six candidates of Generation IV advanced nuclear reactor, which is characterized by remarkable advantages in safety, economics and sustainable development of the fissile resource and proliferation resistance of nuclear energy. A detailed computational model using the Monte Carlo code MCNP5 is set up, in order to study about radical/axis neutronflux and the influences of the reflect thickness or blanket thickness on the conversion ratio (CR) of the Th-U fuel cycle. Main results obtained in this calculation show that: (1) The neutronflux distribution of the graphite-molten zero power reactor core is relatively smooth. (2) CR will increase with the increasing of the thickness of reflector and/or the thickness of breeding region in a certain range and when it exceeds this range CR cannot get increased significantly.

This paper presents an overview of research conducted with the support of the Australian Institute of Nuclear Science and Engineering, at the University of Melbourne, School of Earth Sciences, Radiochemical Neutron Activation Laboratory. The primary objective of this research is to realize the high potential of the platinum group elements (PGE) and gold to the solution of petrogenetic problems, the study of magma generation and magmatic processes in mafic/ultramafic rock suites, as tracers in hydrothermal ore formation. The PGEs (Os, Ru, Ir, Pt, Pd and Rh) are among the least abundant of all elements on earth with unique properties such as high melting points, high electrical and thermal conductivity, high density, strength and toughness as alloys. They exhibit both siderophile and chalcophile characteristics and are valuable tools in providing information about magmatic processes, in particular S-saturation, as well as crystal fractionation trends. Two distinct groups of PGEs are discerned; the IPGEs (Ru, Os, Ir) and the PPGEs (Pt, Pd, Rh, Au) on the basis of their behaviour during fractionation processes. Using chondrite normalized PGE patterns it is possible to distinguish between sulphides that segregated from primitive magmas, such as komatiites, and sulphides which segregated from more fractionated magmas, such as tholeiites. It is critical to the understanding of these processes to be able to analyse key elements, such as the PGE and gold, in the parts per billion to parts per trillion range. Platinum group elements and Au were determined by radiochemical neutron activation analysis using a modified NiS fire-assay preconcentration technique, adapted from procedures first used by Robert, R.V. D. and van Wyk, E. (1975) . Detection limits are generally 0.005-0.01 ppb (Au and Ir), 0.1-0.2 ppb (Pd and Pt), and 0.1-0.5 ppb for Ru. 9 refs.

Neutron scattering techniques have been part of the Australian scientific research community for the past three decades. The High Flux Australian Reactor (HIFAR) is a multi-use facility of modest performance that provides the only neutron source in the country suitable for neutron scattering. The limitations of HIFAR have been recognized and recently a Government initiated inquiry sought to evaluate the future needs of a neutron source. In essence, the inquiry suggested that a delay of several years would enable a number of key issues to be resolved, and therefore a more appropriate decision made. In the meantime, use of the present source is being optimized, and where necessary research is being undertaken at major overseas neutron facilities either on a formal or informal basis. Australia has, at present, a formal agreement with the Rutherford Appleton Laboratory (UK) for access to the spallation source ISIS. Various aspects of neutron scattering have been implemented on HIFAR, including investigations of the structure of biological relevant molecules. One aspect of these investigations will be presented. Preliminary results from a study of the interaction of the immunosuppressant drug, cyclosporin-A, with reconstituted membranes suggest that the hydrophobic drug interdigitated with lipid chains.

Diffusion theory is traditionally applied to nuclear reactor global calculations. Based on the good results generated by the one-dimensional spectral nodal diffusion (SND) method for benchmark problems, we offer the SND method for nuclear reactor global calculations in X,Y geometry. In this method, the continuity equation and Flick law are transverse integrated in each spatial direction leading to a system of two 'one-dimensional' equations coupled by the transverse leakage terms. We then apply the SND method to numerically solve this system with constant approximations for the transverse leakage terms. We perform a spectral analysis to determine the local general solution of each 'one-dimensional' nodal equation with flat approximation for the transverse leakages. We used special auxiliary equations with parameters that are to be determined in order to preserve the analytical general solutions in the numerical algorithm. By considering continuity conditions at the node interfaces and appropriate boundary conditions, we obtain a solvable system of discretized equations involving the node-edge average scalar fluxes at each estimate of the dominant eigenvalue (k{sub eff}) in the outer power iterations. As we considered approximations to the transverse leakages, the SND method is not free of spatial truncation errors. Nevertheless, it generated good results for the typical model problems that we considered. (author)

High-order numerical methods for unstructured grids combine the superior accuracy of high-order spectral or finite difference methods with the geometric flexibility of low-order finite volume or finite element schemes. The Flux Reconstruction (FR) approach unifies various high-order schemes for unstructured grids within a single framework. Additionally, the FR approach exhibits a significant degree of element locality, and is thus able to run efficiently on modern streaming architectures, such as Graphical Processing Units (GPUs). The aforementioned properties of FR mean it offers a promising route to performing affordable, and hence industrially relevant, scale-resolving simulations of hitherto intractable unsteady flows within the vicinity of real-world engineering geometries. In this paper we present PyFR, an open-source Python based framework for solving advection-diffusion type problems on streaming architectures using the FR approach. The framework is designed to solve a range of governing systems on mi...

Compact deuterium-tritium neutron generators emit fast neutrons (14.2 MeV) that have to be thermalized for neutron activation analysis experiments. To maximize thermal neutronflux and minimize epithermal and fast neutronfluxes across the output surface of the neutron generator facility, Monte Carlo calculations (MCNP5; Los Alamos National Laboratory) for different moderator types and widths and collimator and reflector designs have been performed. A thin lead layer close to the neutron generator as neutron multiplier followed by polyethylene moderator and surrounded by a massive lead and nickel collimator and reflector was obtained as the optimum setup.

This report documents the computer code VALE designed to solve multigroup neutronicsproblems with the diffusion theory approximation to neutron transport for a triagonal arrangement of mesh points on planes in two- and three-dimensional geometry. This code parallels the VENTURE neutronics code in the local computation system, making exposure and fuel management capabilities available. It uses and generates interface data files adopted in the cooperative effort sponsored by Reactor Physics RRT Division of the US DOE. The programming in FORTRAN is straightforward, although data is transferred in blocks between auxiliary storage devices and main core, and direct access schemes are used. The size of problems which can be handled is essentially limited only by cost of calculation since the arrays are variably dimensioned. The memory requirement is held down while data transfer during iteration is increased only as necessary with problem size. There is provision for the more common boundary conditions including the repeating boundary, 180/sup 0/ rotational symmetry, and the rotational symmetry conditions for the 30/sup 0/, 60/sup 0/, and 120/sup 0/ triangular grids on planes. A variety of types of problems may be solved: the usual neutronflux eignevalue problem, or a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations. The adjoint problem and fixed source problem may be solved, as well as the dominating higher harmonic, or the importance problem for an arbitrary fixed source.

The flux of fast neutrons produced by CR muons in lead at the depth of 25 mwe is measured. Lead is a common shielding material and neutrons produced in it in muon interactions are unavoidable background component, even in sensitive deep underground experiments. A low background gamma spectrometer, equipped with high purity Ge detector in coincidence with muon detector is used for this purpose. Neutrons are identified by the structure at 692 KeV in the spectrum of delayed coincidences, caused by the neutron inelastic scattering on Ge-72 isotope. Preliminary result for the fast neutron rate is 3.1(5) × 10--4n/cm2 · s.

We demonstrate a novel procedure to calibrate neutron detection systems commonly used in standard neutron radiography. This calibration allows determining the uncertainties due to Poisson-like neutron counting statistics for each individual pixel of a radiographic image. The obtained statistical errors are necessary in order to perform correct quantitative analysis. This fast and convenient method is applied to real data measured at the cold neutron radiography facility ICON at the Paul Scherrer Institute. Moreover, from the results the effective neutronflux at the beam line is determined.

In this technical note, a fractional wave equation for the average neutron motion in nuclear reactor is considered. This representation covers the full spectrum of the average neutron transport behavior, i.e., Fickian and non-Fickian effects. The fractional diffusion model retains the main dynamic characteristics of the neutron motion in which the relaxation time associated with a rapid variation in the neutronflux contains a fractional exponent that can be manipulated to obtain the best representation of the neutron transport phenomena. The detrended fluctuation analysis (DFA) method is presented in this paper to estimate the fractional exponent.

Supermirror coated neutron guides are used at all modern neutron sources for transporting neutrons over large distances. In order to reduce the transmission losses due to multiple internal reflection of neutrons, ballistic neutron guides with linear tapering have been proposed and realized. However, these systems suffer from an inhomogeneous illumination of the sample. Moreover, the flux decreases significantly with increasing distance from the exit of the neutron guide. We propose using elliptically tapered guides that provide a more homogeneous phase space at the sample position as well as a focusing at the sample. Moreover, the design of the guide system is simplified because ellipses are simply defined by their long and short axes. In order to prove the concept we have manufactured a doubly focusing guide and investigated its properties with neutrons. The experiments show that the predicted gains using the program package McStas are realized. We discuss several applications of elliptic guides in various fields of neutron physics.

Supermirror coated neutron guides are used at all modern neutron sources for transporting neutrons over large distances. In order to reduce the transmission losses due to multiple internal reflection of neutrons, ballistic neutron guides with linear tapering have been proposed and realized. However, these systems suffer from an inhomogeneous illumination of the sample. Moreover, the flux decreases significantly with increasing distance from the exit of the neutron guide. We propose using elliptically tapered guides that provide a more homogeneous phase space at the sample position as well as a focusing at the sample. Moreover, the design of the guide system is simplified because ellipses are simply defined by their long and short axes. In order to prove the concept we have manufactured a doubly focusing guide and investigated its properties with neutrons. The experiments show that the predicted gains using the program package McStas are realized. We discuss several applications of elliptic guides in various fields of neutron physics.

We aim at building an accelerator based compact neutron source which would provide a thermal neutronflux on the order of 4E12 n.s-1.cm-2.sr-1. Such brilliance would put compact neutron sources on par with existing medium fluxneutron research reactors. We performed the first neutron production tests on the IPHI proton accelerator at Saclay. The neutronfluxes were measured using gold foil activation and 3He detectors. The measured fluxes were compared with MCNP and GEANT4 Monte Carlo simulations in which the whole experimental setup was modelled. There is a good agreement between the experimental measurements and the Monte-Carlo simulations. The available modelling tools will allow us to optimize the whole Target Moderator Reflector assembly together with the neutron scattering spectrometer geometries.

A case history of the calibration of neutron porosity tools is given in the paper. The calibration of neutron porosity tools is one of the most difficult, complicated, and time consuming tasks in the well logging operations in geophysics. A semi empirical approach to this problem is given in the paper. It is based on the correlation of the tool readings observed in known environments with the apparent neutron parameters sensed by the tools. The apparent neutron parameters are functions of the true neutron parameters of geological formations and of the borehole material, borehole diameter, and the tool position inside the borehole. The true integral neutron transport parameters are obtained by the multigroup diffusion approximation for slowing down of neutrons and by one thermal neutron group for the diffusion. In the latter, the effective neutron temperature is taken into account. The problem of the thermal neutron absorption cross section of rocks is discussed in detail from the point of view of its importance for the well logging results and for the experimental techniques being used.

In non-destructive evaluation with X-rays light elements embedded in dense, heavy (or high-Z) matrices show little contrast and their structural details can hardly be revealed. Neutron radiography, on the other hand, provides a solution for those cases, in particular for hydrogenous materials, owing to the large neutron scattering cross section of hydrogen and uncorrelated dependency of neutron cross section on the atomic number. The majority of neutron imaging experiments at the present time is conducted with static objects mainly due to the limited flux intensity of neutron beamline facilities and sometimes due to the limitations of the detectors. However, some applications require the studies of dynamic phenomena and can now be conducted at several high intensity beamlines such as the recently rebuilt ANTARES beam line at the FRM-II reactor. In this paper we demonstrate the capabilities of time resolved imaging for repetitive processes, where different phases of the process can be imaged simultaneously and...

The Taguchi methodology has proved to be highly efficient to solve inverse problems, in which the values of some parameters of the model must be obtained from the observed data. There are intrinsic mathematical characteristics that make a problem known as inverse. Inverse problems appear in many branches of science, engineering and mathematics. To solve this type of problem, researches have used different techniques. Recently, the use of techniques based on Artificial Intelligence technology is being explored by researches. This paper presents the use of a software tool based on artificial neural networks of generalized regression in the solution of inverse problems with application in high energy physics, specifically in the solution of the problem of neutron spectrometry. To solve this problem we use a software tool developed in the Mat Lab programming environment, which employs a friendly user interface, intuitive and easy to use for the user. This computational tool solves the inverse problem involved in the reconstruction of the neutron spectrum based on measurements made with a Bonner spheres spectrometric system. Introducing this information, the neural network is able to reconstruct the neutron spectrum with high performance and generalization capability. The tool allows that the end user does not require great training or technical knowledge in development and/or use of software, so it facilitates the use of the program for the resolution of inverse problems that are in several areas of knowledge. The techniques of Artificial Intelligence present singular veracity to solve inverse problems, given the characteristics of artificial neural networks and their network topology, therefore, the tool developed has been very useful, since the results generated by the Artificial Neural Network require few time in comparison to other techniques and are correct results comparing them with the actual data of the experiment. (Author)

Earth is connected gravitationally, magnetically and electrically to its heat source - a neutron star that is obscured from view by waste products in the photosphere. Neutron repulsion is like the hot filament in an incandescent light bulb. Excited neutrons are emitted from the solar core and decay into hydrogen that glows in the photosphere like a frosted light bulb. Neutron repulsion was recognized in nuclear rest mass data in 2000 as the overlooked source of energy, the keystone of an arch...

The large neutrino experiments conducted over the last several years at the Los Alamos Neutron Science Center (LANSCE) have provided the opportunity to measure the effects of neutron attenuation in very thick shields. These experiments have featured detectors with active masses of 6 to 150 tons and shield thicknesses ranging from 3000 to 5280 g/cm{sup 2}. An absolute measurement of the high-energy neutronflux was made from the beam stop in a neutrino cave at ninety degrees and nine meters from the beam stop. Differential neutron shielding measurements in iron were also performed, resulting in an attenuation length of 148 g/cm{sup 2}. These measurements allow for the testing of radiation shielding codes for deep penetration problems. The measured flux and attenuation length is compared to calculations using the LAHET Code System (LCS). These codes incorporate biasing techniques, allowing for direct calculation of deep penetration shielding problems. Calculations of the neutron current and attenuation length are presented and compared with measured values. Results from the shielding codes show good agreement with the measured values.

The utilization of hyper-thermal neutrons, which have an energy spectrum with a Maxwellian distribution at a higher temperature than room temperature (300 K), was studied in order to improve the thermal neutronflux distribution at depth in a living body for neutron capture therapy. Simulation calculations were carried out using a Monte Carlo code 'MCNP-V3' in order to investigate the characteristics of hyper-thermal neutrons, i.e. (i) depth dependence of the neutron energy spectrum, and (ii) depth distribution of the reaction rate in a water phantom for materials with 1/v neutron absorption. It is confirmed that hyper-thermal neutron irradiation can improve the thermal neutronflux distribution in the deeper areas in a living body compared with thermal neutron irradiation. When hyper-thermal neutrons with a 3000 K Maxwellian distribution are incident on a body, the reaction rates of 1/v materials such as 14N, 10B etc are about twice that observed for incident thermal neutrons at 300 K, at a depth of 5 cm. The limit of the treatable depth for tumours having 30 ppm 10B is expected to be about 1.5 cm greater by utilizing hyper-thermal neutrons at 3000 K compared with the incidence of thermal neutrons at 300 K.

The accelerator based epithermal neutron source for Boron Neutron Capture Therapy (BNCT) is proposed, created and used in the Budker Institute of Nuclear Physics. In 2014, with the support of the Russian Science Foundation created the BNCT laboratory for the purpose to the end of 2016 get the neutronflux, suitable for BNCT. For getting 3 mA 2.3 MeV proton beam, was created a new type accelerator - tandem accelerator with vacuum isolation. On this moment, we have a stationary proton beam with 2.3 MeV and current 1.75 mA. Generation of neutrons is carried out by dropping proton beam on to lithium target as a result of threshold reaction 7Li(p,n)7Be. Established facility is a unique scientific installation. It provides a generating of neutronflux, including a monochromatic energy neutrons, gamma radiation, alpha-particles and positrons, and may be used by other research groups for carrying out scientific researches. The article describes an accelerator neutron source, presents and discusses the result of experiments and declares future plans.

Kharkov Institute of Physics and Technology (KIPT) of Ukraine has a plan to construct an experimental neutron source facility. The facility has been developed for producing medical isotopes, training young nuclear professionals, supporting the Ukraine nuclear industry, providing capability for performing reactor physics, material research, and basic science experiments. Argonne National Laboratory (ANL) of USA is collaborating with KIPT on developing this facility. A driven subcritical assembly utilizing the KIPT electron accelerator with a target assembly is used to generate the neutron source. The target assembly utilizes tungsten or uranium for neutron production through photonuclear reactions with 100-KW of electron beam power. The neutron source intensity, spectrum, and spatial distribution have been studied to maximize the neutron yield and satisfy different engineering requirements. The subcritical assembly is designed to obtain the highest possible neutronflux intensity with a subcriticality of 0.98. Low enrichment uranium is used for the fuel material because it enhances the neutron source performance. Safety, reliability, and environmental considerations are included in the facility conceptual design. Horizontal neutron channels are incorporated for performing basic research including cold neutron source. This paper describes the conceptual design and summarizes some of the related analyses.

This thesis describes the details on building a new Neutron Interferometry and Optics Facility (NIOFa), the measurement of the incoherent neutron scattering length bi of 3He, and the measurement of the coherent neutron scattering length bc of 4He at National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR). A new monochromatic beamline and facility has been installed at the NCNR devoted to neutron interferometry in the research areas of spin control, spin manipulation, quantum mechanics, quantum information science, spintronics, and material science. This facility is possible in part because of advances in decoherence free subspace interferometer designs that have demonstrated consistent contrast in the presence of vibrational noise; a major environmental constraint that has prevented neutron interferometry from being applied at other neutron facilities. This new facility, NIOFa, is located in the guide hall of the NCNR upstream of the existing Neutron Interferometry and Optics Facility (NIOF) and has several advantages over the NIOF including higher incident flux, better neutron polarization, and increased accessibility. The measurement of the incoherent neutron scattering length bi of 3He was done using a (220) single silicon crystal skew symmetric interferometer. This experiment requires both a polarized beam and a polarized target. We report bi = -2.35 +/- 0.014 (stat.) +/- 0.014 (syst.). This experiment is a revision of the previous experiment which was done in 2008, and partially explains the non-zero phase shift seen in 2008 experiment even if target cell was completely unpolarized. The measurement of the coherent neutron scattering length b c of the 4He was done using a (111) single silicon crystal interferometer. The neutron interferometry and optics facility at NIST had been used previously to determine the coherent scattering lengths for n- 1H, n-2H, and n-3He to less than 1% relative uncertainty. We report bc of the 4He

It is present in schematic way, the origin of the neutrons; their interaction with matter, until its application in the field of dosimetry. It describes some measuring instruments based on thermoluminescence dosimetry, some activation detectors and trace detectors. Finally, it summarizes the work in neutron dosimetry have been carried out at the Autonomous University of Barcelona. (Author)

The CANDU 9 480/SEU core uses 0.9% SEU (Slightly Enriched Uranium) fuel. The use f SEU fuel enables the reactor to increase the radial power form factor from 0.865, which is typical in current natural uranium CANDU reactors, to 0.97 in the nominal CANDU 9 480/SEU core. The difference is a 12% increase in reactor power. An additional 5% increase can be achieved due to a reduced refuelling ripple. The channel power limits were also increased by 3% for a total reactor power increase of 20%. This report describes the calculation of neutronflux distributions in the CANDU 9 480/SEU core under conditions specified by the C and I engineers. The RFSP code was used to calculate of neutronflux shapes for ROP analysis. Detailed flux values at numerous potential detector sites were calculated for each flux shape. (author). 6 tabs., 70 figs., 4 refs.

A fast neutron spectrometer consisting of segmented plastic scintillator and He-3 proportional counters was constructed for the measurement of neutrons in the energy range 1 MeV to 200 MeV. We discuss its design, principles of operation, and the method of analysis. The detector is capable of observing very low neutronfluxes in the presence of ambient gamma background and does not require scintillator pulse-shape discrimination The spectrometer was characterized for energy resolution and efficiency in fast neutron fields of 2.5 MeV, 14 MeV, and fission spectrum neutrons, and the results are compared with Monte Carlo simulations. Measurements of the fast neutronflux and energy response at 120 m above sea-level (39.130 deg. N, 77.218 deg. W) and at a depth of 560 m in a limestone mine are presented. Finally, the design of a spectrometer with improved sensitivity and energy resolution is discussed.

6 statistically significant areas in which it was recorded a lower value of the flow of epithermal neutrons was found in the polar regions of the moon according to LEND: 5 areas are located in the south polar region (the area NSR S1 - 5 [1]) and one (area NSR N1[1]) to the north. One of these areas - the area NSR S5 - is located in the landing sector Luna - Globe mission [2], the launch of which is planned by Russian Space Agency in 2018. In this paper, we investigated the temperature regime, illumination conditions and the possibility of the existence of deposits of volatile compounds in this area. To study we selected a number of substances was detected in the LCROSS impact site in the crater Cabeus, particularly compounds such as H2O, CO2, SO2, CH3OH, NH3, C2H4, H2S, CH4 · 5.75H2O and CO · 5.75H2O [3]. We divided the area of NSR S5 spot into a grid with a number of elements. Step in longitude grid was 0.15 degrees, a step in latitude 0.05 degrees. The total number of the elements of the area of the crater is 36000. The height, slope and orientation of each element were calculated based on a LOLA DEM [4] using an algorithm described in [5]. Our results show that the compounds of deposits such as H2O, CO2, SO2, CH3OH, NH3, C2H4, H2S, CH4 · 5.75H2O and CO · 5.75H2O may exist in NSR S5. Thus, the local suppression the epithermal neutronflux in this region may be due to the presence of hydrogen-containing compounds deposits. [1] Mitrofanov et al. (2012) JGR 117, E003956 [2] Ivanov et al. (2014) Solar System Res. 48, 391 - 402 [3] Colaprete et al. (2010) Science 330, 463-468 [4] http://wwwpds.wustl.edu/ [5] Zevenbergen, L.W., Thorne (1987) Earth Surface Processes and Landforms 12(1), 47-56.

Parameter studies, design calculations and neutronic performance measurements have been completed for a new thermal neutron beamline constructed for neutron capture therapy cell and small-animal radiobiology studies at the University of Missouri Research Reactor. The beamline features the use of single-crystal silicon and bismuth sections for neutron filtering and for reduction of incident gamma radiation. The computational models used for the final beam design and performance evaluation are based on coupled discrete-ordinates and Monte Carlo techniques that permit detailed modeling of the neutron transmission properties of the filtering crystals with very few approximations. Validation protocols based on neutron activation spectrometry measurements and rigorous least-square adjustment techniques show that the beam produces a neutron spectrum that has the anticipated level of thermal neutronflux and a somewhat higher than expected, but radio-biologically insignificant, epithermal neutronflux component. (authors)

A series of measurements has been completed that provides a benchmark for Monte Carlo simulations related to an algorithm for explosives detection using active neutron interrogation. The original simulations used in algorithm development, based on land-sea cargo container screening, have been adapted to model active neutron interrogation of smaller targets. These smaller-scale measurements are easily accomplished in a laboratory environment. Benchmarking measurements were completed using a D-D neutron generator, two neutron detectors, as well as a variety of scatter media including the explosives surrogate melamine (C{sub 3}H{sub 6}N{sub 6}). Measurements included 90 Degree-Sign , 120 Degree-Sign , or 150 Degree-Sign neutron scatter geometries and variations in source-detector shielding, target presence, and target identity. Comparisons of measured and simulated neutronfluxes were similar, with correlation coefficients greater than 0.7. The simulated detector responses also matched very closely with the measured photon and neutron pulse height distributions, with correlation coefficients exceeding 0.9. The experiments and simulations also provided insight into potential application of the new method to the problem of explosives detection in small objects such as luggage and small packages.

In principle, leading neutrons produced in photoproduction and deep-inelastic scattering at HERA have the potential to determine the pion structure function, the neutron absorptive cross section and the form of the pion flux. To explore this potential we compare theoretical predictions for the x_L and p_t spectra of leading neutrons, and the Q^2 dependence of the cross section, with the existing ZEUS data.

real Banach space which is also uniformly smooth using the properties of generalized f-projection operator. Using this result, we discuss strong convergence theorem concerning general H-monotone mappings and system of generalized mixed equilibrium problems in Banach spaces. Our results extend many known recent results in the literature.

Next to fission and fusion, spallation is an efficient process for releasing neutrons from nuclei. Unlike the other two reactions, it is an endothermal process and can, therefore, not be used per se in energy generation. In order to sustain a spallation reaction, an energetic beam of particles, most commonly protons, must be supplied onto a heavy target. Spallation can, however, play an important role as a source of neutrons whose flux can be easily controlled via the driving beam. Up to a few GeV of energy, the neutron production is roughly proportional to the beam power. Although sophisticated Monte Carlo codes exist to compute all aspects of a spallation facility, many features can be understood on the basis of simple physics arguments. Technically a spallation facility is very demanding, not only because a reliable and economic accelerator of high power is needed to drive the reaction, but also, and in particular, because high levels of radiation and heat are generated in the target which are difficult to cope with. Radiation effects in a spallation environment are different from those commonly encountered in a reactor and are probably even more temperature dependent than the latter because of the high gas production rate. A commonly favored solution is the use of molten heavy metal targets. While radiation damage is not a problem in this case, except for the container, a number of other issues are discussed. (author)

The neutron imaging facility ANTARES, operated by the Technische Universität München, is located at the cold neutron beam port SR-4a. Based on a pinhole camera principle with a variable collimator located close to the beam port, the facility provides the possibility for flexible use in high resolution and high flux imaging.

The recent developments in scientific complementary metal oxide semiconductor (sCMOS) detector technology allow for imaging of relevant processes with very high temporal resolution with practically negligible readout time. However, it is neutron intensity that limits the high temporal resolution neutron imaging. In order to partially overcome the neutron intensity problem for the high temporal resolution imaging, a parabolic neutron focussing guide was utilized in the test arrangement and placed upstream the detector in such a manner that the focal point of the guide was positioned slightly behind the scintillator screen. In such a test arrangement, the neutronflux can be increased locally by about one order of magnitude, albeit with the reduced spatial resolution due to the increased divergence of the neutron beam. In a pilot test application, an in-situ titration system allowing for a remote delivery of well-defined volumes of liquids onto the sample stage was utilized. The process of droplets of water (H2O) falling into the container filled with heavy water (D2O) and the subsequent process of the interaction and mixing of the two liquids were imaged with temporal resolution of 0.01 s. •Combination of neutron focussing device and use of sCMOS detector allows for very high temporal resolution neutron imaging to be achieved (albeit with reduced spatial resolution and field of view).•In-situ neutron imaging titration device for liquid interaction experiments.•Interaction of otherwise indiscernible liquids (H2O and D2O) visualized using neutron radiography with 0.01 s temporal resolution.

CARR will provide with maximal neutronflux in Asia,the third of the world.By using the high quality neutron beam and the advanced international experience,Prompt Gamma Neutron Activation Analysis(PGNAA)facility will be setup at high level.PGNAA on CARR will promote the development of nuclear analysis technology and improve Chinese status in the nuclear analysis field.

High-order numerical methods for unstructured grids combine the superior accuracy of high-order spectral or finite difference methods with the geometric flexibility of low-order finite volume or finite element schemes. The Flux Reconstruction (FR) approach unifies various high-order schemes for unstructured grids within a single framework. Additionally, the FR approach exhibits a significant degree of element locality, and is thus able to run efficiently on modern streaming architectures, such as Graphical Processing Units (GPUs). The aforementioned properties of FR mean it offers a promising route to performing affordable, and hence industrially relevant, scale-resolving simulations of hitherto intractable unsteady flows within the vicinity of real-world engineering geometries. In this paper we present PyFR, an open-source Python based framework for solving advection-diffusion type problems on streaming architectures using the FR approach. The framework is designed to solve a range of governing systems on mixed unstructured grids containing various element types. It is also designed to target a range of hardware platforms via use of an in-built domain specific language based on the Mako templating engine. The current release of PyFR is able to solve the compressible Euler and Navier-Stokes equations on grids of quadrilateral and triangular elements in two dimensions, and hexahedral elements in three dimensions, targeting clusters of CPUs, and NVIDIA GPUs. Results are presented for various benchmark flow problems, single-node performance is discussed, and scalability of the code is demonstrated on up to 104 NVIDIA M2090 GPUs. The software is freely available under a 3-Clause New Style BSD license (see www.pyfr.org). Catalogue identifier: AETY_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AETY_v1_0.html Program obtainable from: CPC Program Library, Queen’s University, Belfast, N. Ireland Licensing provisions: New style BSD license No. of lines in

We propose a method for observation of the quasistationary states of neutrons localized near a curved mirror surface. The bounding effective well is formed by the centrifugal potential and the mirror Fermi potential. This phenomenon is an example of an exactly solvable “quantum bouncer” problem that can be studied experimentally. It could provide a promising tool for studying fundamental neutron-matter interactions, as well as quantum neutron optics and surface physics effects. We develop a formalism that describes quantitatively the neutron motion near the mirror surface. The effects of mirror roughness are taken into account.

Neutron radiography and its related two-dimensional (2D) neutron imaging techniques have been established as invaluable nondestructive inspection methods and quantitative measurement tools. They have been used in a wide variety of applications ranging from inspection of aircraft engine turbine blades to study of two-phase fluid flow in operating proton exchange membrane fuel cells. Neutron radiography is similar to X-ray radiography in that the method produces a 2D attenuation map of neutron radiation that has penetrated the object being examined. However, the images produced differ and are often complementary due to the differences between X-ray and neutron interaction mechanisms. The uses and types of 2D neutron imaging have expanded over the past 15 years as a result of advances in imaging technology and improvements in neutron generators/sources and computers. Still, high-intensity sources such as those from reactors and spallation neutron sources, together with conventional film radiography, remain the mainstay of high-resolution, large field-of-view neutron imaging. This chapter presents a summary of the history, methods, and related variations of neutron radiography techniques.

The progress made on investigations of neutron radiography since October 1960 is discussed. The problems of the production of the neutron image will be discussed. The emphasis will be on the characteristics of many of the photographic imaging methods which can be used for neutron radiography. (auth)

The natural neutron background influences the maximum achievable sensitivity in most deep underground nuclear, astroparticle and double-beta decay physics experiments. Reliable neutronflux numbers are an important ingredient in the design of the shielding of new large-scale experiments as well as in the analysis of experimental data. Using a portable setup of He-3 counters we measured the thermal neutronflux at the Kimballton Underground Research Facility, the Soudan Underground Laboratory, on the 4100 ft and the 4850 ft levels of the Sanford Underground Research Facility, at the Waste Isolation Pilot Plant and at the Gran Sasso National Laboratory. Absolute neutronfluxes at these laboratories are presented.

The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. High-resolution gamma-ray spectrometers are used in these applications to measure the spectrum of the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used simulation tool for this type of problem, but computational times can be prohibitively long. This work explores the use of multi-group deterministic methods for the simulation of coupled neutron-photon problems. The main purpose of this work is to benchmark several problems modeled with RADSAT and MCNP to experimental data. Additionally, the cross section libraries for RADSAT are updated to include ENDF/B-VII cross sections. Preliminary findings show promising results when compared to MCNP and experimental data, but also areas where additional inquiry and testing are needed. The potential benefits and shortcomings of the multi-group-based approach are discussed in terms of accuracy and computational efficiency.

The cosmogenic neutron yield Yn characterizes the ability of matter to produce neutrons under the effect of cosmic ray muons with spectrum and average energy corresponding to an observation depth. The yield is the basic characteristic of cosmogenic neutrons. The neutron production rate and neutronflux both are derivatives of the yield. The constancy of the exponents α and β in the known dependencies of the yield on energy Yn∝Eμα and the atomic weight Yn∝Aβ allows one to combine these dependencies in a single formula and to connect the yield with muon energy loss in matter. As a result, the phenomenological formulas for the yields of muon-induced charged pions and neutrons can be obtained. These expressions both are associated with nuclear loss of the ultrarelativistic muons, which provides the main contribution to the total neutron yield. The total yield can be described by a universal formula, which is the best fit of the experimental data.

The RCP01 Monte Carlo program is used to analyze many geometries of interest in nuclear design and analysis of light water moderated reactors such as the core in its pressure vessel with complex piping arrangement, fuel storage arrays, shipping and container arrangements, and neutron detector configurations. Written in FORTRAN and in use on a variety of computers, it is capable of estimating steady state neutron or photon reaction rates and neutron multiplication factors. The energy range covered in neutron calculations is that relevant to the fission process and subsequent slowing-down and thermalization, i.e., 20 MeV to 0 eV. The same energy range is covered for photon calculations.

For the research of CCD neutron radiography, a neutron collimator was designed based on the exit of thermal neutron of the Boron Neutron Capture Therapy (BNCT) reactor. Based on the Geant4 simulations, the preliminary choice of the size of the collimator was determined. The materials were selected according to the literature data. Then, a collimator was constructed and tested on site. The results of experiment and simulation show that the thermal neutronflux at the end of the neutron collimator is greater than 1.0×106 n/cm2/s, the maximum collimation ratio (L/D) is 58, the Cd-ratio(Mn) is 160 and the diameter of collimator end is 10 cm. This neutron collimator is considered to be applicable for neutron radiography.

This book provides a systematic and comprehensive introduction to fusion neutronics, covering all key topics from the fundamental theories and methodologies, as well as a wide range of fusion system designs and experiments. It is the first-ever book focusing on the subject of fusion neutronics research. Compared with other nuclear devices such as fission reactors and accelerators, fusion systems are normally characterized by their complex geometry and nuclear physics, which entail new challenges for neutronics such as complicated modeling, deep penetration, low simulation efficiency, multi-physics coupling, etc. The book focuses on the neutronics characteristics of fusion systems and introduces a series of theories and methodologies that were developed to address the challenges of fusion neutronics, and which have since been widely applied all over the world. Further, it introduces readers to neutronics design’s unique principles and procedures, experimental methodologies and technologies for fusion systems...

Oak Ridge National Laboratory (ORNL) is leveraging decades of experience in neutron characterization of advanced materials together with resources such as the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR) shown in Fig. 1 to solve challenging problems in additive manufacturing (AM). Additive manufacturing, or three-dimensional (3-D) printing, is a rapidly maturing technology wherein components are built by selectively adding feedstock material at locations specified by a computer model. The majority of these technologies use thermally driven phase change mechanisms to convert the feedstock into functioning material. As the molten material cools and solidifies, the component is subjected to significant thermal gradients, generating significant internal stresses throughout the part (Fig. 2). As layers are added, inherent residual stresses cause warping and distortions that lead to geometrical differences between the final part and the original computer generated design. This effect also limits geometries that can be fabricated using AM, such as thin-walled, high-aspect- ratio, and overhanging structures. Distortion may be minimized by intelligent toolpath planning or strategic placement of support structures, but these approaches are not well understood and often "Edisonian" in nature. Residual stresses can also impact component performance during operation. For example, in a thermally cycled environment such as a high-pressure turbine engine, residual stresses can cause components to distort unpredictably. Different thermal treatments on as-fabricated AM components have been used to minimize residual stress, but components still retain a nonhomogeneous stress state and/or demonstrate a relaxation-derived geometric distortion. Industry, federal laboratory, and university collaboration is needed to address these challenges and enable the U.S. to compete in the global market. Work is currently being conducted on AM technologies at the ORNL

Due to the high neutron yield and the large plasma size many ITER plasma parameters such as fusion power, power density, ion temperature, fast ion energy and their spatial distributions in the plasma core can be measured well by various neutron diagnostics. Neutron diagnostic systems under consideration and development for ITER include radial and vertical neutron cameras (RNC and VNC), internal and external neutronflux monitors (NFMs), neutron activation systems and neutron spectrometers. The two-dimensional neutron source strength and spectral measurements can be provided by the combined RNC and VNC. The NFMs need to meet the ITER requirement of time-resolved measurements of the neutron source strength and can provide the signals necessary for real-time control of the ITER fusion power. Compact and high throughput neutron spectrometers are under development. A concept for the absolute calibration of neutron diagnostic systems is proposed. The development, testing in existing experiments and the engineering integration of all neutron diagnostic systems into ITER are in progress and the main results are presented.

A new, algebraic, method is applied to calculation of neutron albedo from substance to check the claim that use of ultradispersive fuel and moderator of an active core can help to gain in size and mass of the reactor. In a model of isotropic distribution of incident and reflected neutrons it is shown that coherent scattering on separate grains in the case of thermal neutrons increases transport cross section negligibly, however it decreases albedo from a wall of finite thickness because of decrease of substance density. A visible increase of albedo takes place only for neutrons with wave length of the order of the size of a single grain.

Design studies relating to an in vivo prompt capture neutron activation analysis facility measuring total body nitrogen are presented. The basis of the design is a beryllium-graphite neutron collimator and reflector configuration for (alpha, n) type radionuclide neutron sources (238PuBe or 241AmBe), so as to reflect leaking, or out-scattered, neutrons towards the subject. This improves the ratio of thermal neutronflux to dose and the spatial distribution of thermal flux achieved with these sources, whilst retaining their advantage of long half-lives as compared to 252Cf based systems. The common problem of high count-rate at the detector, and therefore high nitrogen region of interest background due to pile-up, is decreased by using a set of smaller (5.1 cm diameter x 10.2 cm long) NaI(Tl) detectors instead of large ones. The facility described presents a relative error of nitrogen measurement of 3.6% and a nitrogen to background ratio of 2.3 for 0.45 mSv skin dose (assuming ten 5.1 cm x 10.2 cm NaI(Tl) detectors).

Design studies relating to an in vivo prompt capture neutron activation analysis facility measuring total body nitrogen are presented. The basis of the design is a beryllium-graphite neutron collimator and reflector configuration for ({alpha}, n) type radionuclide neutron sources ({sup 238}PuBe or {sup 241}AmBe), so as to reflect leaking, or out-scattered, neutrons towards the subject. This improves the ratio of thermal neutronflux to dose and the spatial distribution of thermal flux achieved with these sources, whilst retaining their advantage of long half-lives as compared to {sup 252}Cf based systems. The common problem of high count-rate at the detector, and therefore high nitrogen region of interest background due to pile-up, is decreased by using a set of smaller (5.1 cm diameter x 10.2 cm long) NaI(Tl) detectors instead of large ones. The facility described presents a relative error of nitrogen measurement of 3.6% and a nitrogen to background ratio of 2.3 for 0.45 mSv skin dose (assuming ten 5.1 cm x 10.2 cm NaI(Tl) detectors). (author).

Circulardiscs of Li2SiO3 doped with europium were prepared and a new activation procedure for the neutron dose estimation in a breeder blanket of fusion reactor is described. The amount of europium in the disc was determined by neutron activation analysis (NAA) using an isotopic neutron source. The average neutron absorption cross section for the reaction was calculated using neutron distribution of the Am-Be source and available neutron absorption cross section data for the (151)Eu(n,γ)(152m)Eu reaction, which was used for estimation of europium in the pallet. The cross section of the elements varies with neutron energy, and the flux of the neutrons in each energy range seen by the nuclei under investigation also varies. Neutron distribution spectrum of the Am-Be source was worked out prior to NAA and the effective fractional flux for the nuclear reaction considered for the flux estimation was also determined.

We used the Monte Carlo program MCNP to calculate the neutron and gamma fluxes on a fast neutron irradiation facility being installed on the Portuguese Research Reactor (RPI). The purpose of this facility is to provide a fast neutron beam for irradiation of electronic circuits. The gamma dose should be minimized. This is achieved by placing a lead shield preceded by a thin layer of boral. A fast neutronflux of the order of 109 n/cm2s is expected at the exit of the tube, while the gamma radiation is kept below 20 Gy/h. We will present results of the neutron and gamma doses for several locations along the tube and different thickness of the lead shield. We found that the neutron beam is very collimated at the end of the tube with a dominant component on the fast region.

This document describes the initial global 2-D shielding analyses for the Advanced Neutron Source (ANS) reactor, the D{sub 2}O reflector, the reflector vessel, and the first 200 mm of light water beyond the reflector vessel. Flux files generated here will later serve as source terms in subsequent shielding analyses. In addition to reporting fluxes and other data at key points of interest, a major objective of this report was to document how these analyses were performed, the phenomena that were included, and checks that were made to verify that these phenomena were properly modeled. In these shielding analyses, the fixed neutron source distribution in the core was based on the `lifetime-averaged` spatial power distribution. Secondary gamma production cross sections in the fuel were modified so as to account intrinsically for delayed fission gammas in the fuel as well as prompt fission gammas. In and near the fuel, this increased the low-energy gamma fluxes by 50 to 250%, but out near the reflector vessel, these same fluxes changed by only a few percent. Sensitivity studies with respect to mesh size were performed, and a new 2-D mesh distribution developed after some problems were discovered with respect to the use of numerous elongated mesh cells in the reflector. All of the shielding analyses were performed sing the ANSL-V 39n/44g coupled library with 25 thermal neutron groups in order to obtain a rigorous representation of the thermal neutron spectrum throughout the reflector. Because of upscatter in the heavy water, convergence was very slow. Ultimately, the fission cross section in the various materials had to be artificially modified in order to solve this fixed source problem as an eigenvalue problem and invoke the Vondy error-mode extrapolation technique which greatly accelerated convergence in the large 2-D RZ DORT analyses. While this was quite effective, 150 outer iterations (over energy) were still required.

This document describes the initial global 2-D shielding analyses for the Advanced Neutron Source (ANS) reactor, the D{sub 2}O reflector, the reflector vessel, and the first 200 mm of light water beyond the reflector vessel. Flux files generated here will later serve as source terms in subsequent shielding analyses. In addition to reporting fluxes and other data at key points of interest, a major objective of this report was to document how these analyses were performed, the phenomena that were included, and checks that were made to verify that these phenomena were properly modeled. In these shielding analyses, the fixed neutron source distribution in the core was based on the `lifetime-averaged` spatial power distribution. Secondary gamma production cross sections in the fuel were modified so as to account intrinsically for delayed fission gammas in the fuel as well as prompt fission gammas. In and near the fuel, this increased the low-energy gamma fluxes by 50 to 250%, but out near the reflector vessel, these same fluxes changed by only a few percent. Sensitivity studies with respect to mesh size were performed, and a new 2-D mesh distribution developed after some problems were discovered with respect to the use of numerous elongated mesh cells in the reflector. All of the shielding analyses were performed sing the ANSL-V 39n/44g coupled library with 25 thermal neutron groups in order to obtain a rigorous representation of the thermal neutron spectrum throughout the reflector. Because of upscatter in the heavy water, convergence was very slow. Ultimately, the fission cross section in the various materials had to be artificially modified in order to solve this fixed source problem as an eigenvalue problem and invoke the Vondy error-mode extrapolation technique which greatly accelerated convergence in the large 2-D RZ DORT analyses. While this was quite effective, 150 outer iterations (over energy) were still required.

Neutron radiography has the ability to differentiate between gas and liquid in two-phase flow due both to the density difference and the high neutron scattering probability of hydrogen. Previous studies have used dynamic neutron radiography -- in both real-time and high-speed -- for air-water, steam-water and gas-liquid metal two-phase flow measurements. Radiography with thermal neutrons is straightforward and efficient as thermal neutrons are easier to detect with relatively higher efficiency and can be easily extracted from nuclear reactor beam ports. The quality of images obtained using neutron radiography and the imaging speed depend on the neutron beam intensity at the imaging plane. A high quality neutron beam, with thermal neutron intensity greater than 3.0x 10 6 n/cm2-s and a collimation ratio greater than 100 at the imaging plane, is required for effective dynamic neutron radiography up to 2000 frames per second. The primary objectives of this work are: (1) to optimize a neutron radiography facility for dynamic neutron radiography applications and (2) to investigate a new technique for three-dimensional neutron radiography using information obtained from neutron scattering. In this work, neutron transport analysis and experimental validation of a dynamic neutron radiography facility is studied with consideration of real-time and high-speed neutron radiography requirements. A beam port based dynamic neutron radiography facility, for a target thermal neutronflux of 1.0x107 n/cm2-s, has been analyzed, constructed and experimentally verified at the McMaster Nuclear Reactor. The neutron source strength at the beam tube entrance is evaluated experimentally by measuring the thermal and fast neutronfluxes using copper activation flux-mapping technique. The development of different facility components, such as beam tube liner, gamma ray filter, beam shutter and biological shield, is achieved analytically using neutron attenuation and divergence theories. Monte

In 2011-2012, an experimental campaign with a significant yield of fusion neutrons was carried out on the JET tokamak. During this campaign the facility was equipped with two diamond detectors based on natural and artificial CVD diamond. These detectors were designed and manufactured in State Research Center of Russian Federation TRINITI. The detectors measure the flux of fast neutrons with energies above 0.2 MeV. They have been installed in the torus hall and the distance from the center of plasma was about 3 m. For some of the JET pulses in this experiment, the neutronflux density corresponded to the operational conditions in collimator channels of ITER Vertical Neutron Camera. The main objective of diamond monitors was the measurement of total fast neutronflux at the detector location and the estimation of the JET total neutron yield. The detectors operate as threshold counters. Additionally a spectrometric measurement channel has been configured that allowed us to distinguish various energy components of the neutron spectrum. In this paper we describe the neutron signal measuring and calibration procedure of the diamond detector. Fluxes of DD and DT neutrons at the detector location were measured. It is shown that the signals of total neutron yield measured by the diamond detector correlate with signals measured by the main JET neutron diagnostic based on fission chambers with high accuracy. This experiment can be considered as a successful test of diamond detectors in ITER-like conditions.

Measurement of the flux and energy spectrum of the protons resulting from the decay of solar flare neutrons gives unique information on the spectrum of neutrons from 5 to 200 MeV. Neutrons from three flares have been observed in this manner during solar cycle 21. The use of the decay protons to determine neutron energy spectra is reviewed, and new and definitive energy spectra are presented for the two large flares on June 3, 1982 and April 25, 1984.

The results of TRANSP simulations of neutron count rate profiles measured by a collimated neutronflux monitor-neutron camera (NC)—for different plasma scenarios on MAST are reported. In addition, the effect of various plasma parameters on neutron emission is studied by means of TRANSP simulation. The fast ion redistribution and losses due to fishbone modes, which belong to a wider category of energetic particle modes, are observed by the NC and modelled in TRANSP.

A neutron irradiation facility was designed and built in order to establish a procedure for calibrating neutron monitors and dosemeters. A 185 GBq {sup 241} Am Be source of known is used as a reference source. The irradiation facility using this source in the air provides neutron dose rates between 9 nSv s{sup -1} and 0,5 {sup {mu}}Sv s{sup -1}. A calibrated 50 nSv s{sup -1} thermal neutron field is obtained by using a specially designed paraffin block in conjunction with the {sup 241} Am Be source. A Bonner multisphere spectrometer was calibrated, using a procedure based on three methods proposed by international standards. The unfold {sup 241} Am Be neutron spectrum was determined from the Bonner spheres data and resulted in a good agreement with expected values for fluence rate, dose rate and mean energy. A dosimetric system based on the electrochemical etching of CR-39 was developed for personal dosimetry. The dosemeter badge using a (n,{alpha}) converter, the etching chamber and high frequency power supply were designed and built specially for this project. The electrochemical etching (ECE) parameters used were: a 6N KOH solution, 59 deg C, 20 kV{sub pp} cm{sup -1}, 2,0 kHz, 3 hours of ECE for thermal and intermediate neutrons and 6 hours for fast neutrons. The calibration factors for thermal, intermediate and fast neutrons were determined for this personal dosemeter. The sensitivities determined for the developed dosimetric system were (1,46{+-} 0,09) 10{sup 4} tracks cm{sup -2} mSv{sup -1} for thermal neutrons, (9{+-}3) 10{sup 2} tracks cm{sup -2} mSV{sup -1} for intermediate neutrons and (26{+-}4) tracks cm{sup -2} mSv{sup -1} for fast neutrons. The lower and upper limits of detection were respectively 0,002 mSv and 0,6 mSv for thermal neutrons, 0,04 mSv and 8 mSv for intermediate neutrons and 1 mSv and 12 mSv for fast neutrons. In view of the 1990`s ICRP recommendations, it is possible to conclude that the personal dosemeter described in this work is

The neutron time-of-flight facility n_TOF at CERN, Switzerland, operational since 2001, delivers neutrons using the Proton Synchrotron (PS) 20 GeV/c proton beam impinging on a lead spallation target. The facility combines a very high instantaneous neutronflux, an excellent time of flight resolution due to the distance between the experimental area and the production target (185 meters), a low intrinsic background and a wide range of neutron energies, from thermal to GeV neutrons. These characteristics provide a unique possibility to perform neutron-induced capture and fission cross-section measurements for applications in nuclear astrophysics and in nuclear reactor technology.

The MEGAPIE project is one of the key experiments towards the feasibility of Accelerator Driven Systems. On-line operation and post-irradiation analysis will provide the scientific community with unique data on the behavior of a liquid spallation target under realistic irradiation conditions. A good neutronics performance of such a target is of primary importance towards an intense neutron source, where an extended liquid metal loop requires some dedicated verifications related to the delayed neutron activity of the irradiated PbBi. In this paper we report on the experimental characterization of the MEGAPIE neutronics in terms of the prompt neutron (PN) flux inside the target and the delayed neutron (DN) flux on the top of it. For the PN measurements, a complex detector, made of 8 microscopic fission chambers, has been built and installed in the central part of the target to measure the absolute neutronflux and its spatial distribution. Moreover, integral information on the neutron energy distribution as a f...

The Frankfurt neutron source of Stern Gerlach Zentrum FRANZ is currently under construction at the University of Frankfurt. At FRANZ, a high intensity neutron beam in the keV energy region will be produced by bombarding a $^7$Li target with a proton beam of several mA. These unprecedented high neutronfluxes will allow a number of neutron induced cross section measurements for the first time. Measurements can be performed by the time-of-flight and by the activation technique.

In Part I of this paper, we described the use of the computer code SAIPS in neutron spectrum unfolding. Here in Part II, we present our experimental work carried out to study the shape of the neutron spectrum in different experimental channels of a 5 MW light-water cooled and moderated research reactor. The spectral neutronflux was determined using various fission foils (placed in close contact with mica track detectors) and activation detectors. From the measured activities, the neutron spectrum was unfolded by SAIPS. (author).

Magnetic properties of superconductors with peak effect were investigated both experimentally and theoretically in frames of the critical state model which incorporates the flux jump instability criterion. Theoretical analyses show some “forbidden” band for flux jumps on the magnetic field axis. Features of H-T diagrams of instability of superconductors with peak effect are discussed.

A fast neutron spectrometer consisting of segmented plastic scintillator and 3He proportional counters was constructed for the measurement of neutrons in the energy range 1-200 MeV. We discuss its design, principles of operation, and the method of analysis. The detector is capable of observing very low neutronfluxes in the presence of ambient gamma background and does not require scintillator pulse-shape discrimination. The spectrometer was characterized for its energy response in fast neutron fields of 2.5 MeV and 14 MeV, and the results are compared with Monte Carlo simulations. Measurements of the fast neutronflux and energy response at 120 m above sea-level (39.130°N, 77.218°W) and at a depth of 560 m in a limestone mine are presented. Finally, the design of a spectrometer with improved sensitivity and energy resolution is discussed.

A fast neutron spectrometer consisting of segmented plastic scintillator and {sup 3}He proportional counters was constructed for the measurement of neutrons in the energy range 1–200 MeV. We discuss its design, principles of operation, and the method of analysis. The detector is capable of observing very low neutronfluxes in the presence of ambient gamma background and does not require scintillator pulse-shape discrimination. The spectrometer was characterized for its energy response in fast neutron fields of 2.5 MeV and 14 MeV, and the results are compared with Monte Carlo simulations. Measurements of the fast neutronflux and energy response at 120 m above sea-level (39.130°N, 77.218°W) and at a depth of 560 m in a limestone mine are presented. Finally, the design of a spectrometer with improved sensitivity and energy resolution is discussed.

In the framework of the Neutronic and Nuclear Assessment Task Group of the MEGAPIE experiment we measured the delayed neutron (DN) flux at the top of the target. The measurement was proposed mainly for radioprotection purposes since the DN flux at the top of the target has been estimated to be of the same order of magnitude as the prompt neutronflux. Given the strong model-dependence of DN predictions, the measurement of DN contribution to the total neutron activity at the top of the target was thus desired. Moreover, this measurement is complementary to the DN experiments performed at PNPI (Gatchina) on solid lead and bismuth targets. The DN measurement at MEGAPIE was performed during the start-up phase of the target. In this paper we present a detailed description of the experimental setup and some preliminary results on decay spectra. (authors)

In the framework of the Neutronic and Nuclear Assessment Task Group of the MEGAPIE experiment we measured the delayed neutron (DN) flux at the top of the target. The measurement was proposed mainly for radioprotection purposes since the DN flux at the top of the target has been estimated to be of the same order of magnitude as the prompt neutronflux. Given the strong model-dependence of DN predictions, the measurement of DN contribution to the total neutron activity at the top of the target was thus desired. Moreover, this measurement is complementary to the DN experiments performed at PNPI (Gatchina) on solid lead and bismuth targets. The DN measurement at MEGAPIE was performed during the start-up phase of the target. In this paper we present a detailed description of the experimental setup and some preliminary results on decay spectra.

The use of the U-120 Cyclotron of the IFIN-HH allowed to perform a testing bench with fast neutrons in order to determine the correction factors of the doseratemeters dedicated to neutron measurement. This paper deals with researchers performed in order to develop the irradiation facility testing the fast neutronsflux generated at the Cyclotron. This facility is presented, together with the results obtain in determining the correction factor for a doseratemeter dedicated to the neutron dose equivalent rate measurement.

Due to the variability of neutron spectrum within the same environment, it is essential that the spectral distribution as function of energy to be characterized. To perform this task, the neutron spectrometer has a primary role in determining the neutronflux ({Phi}{sub E}(E)). Precise information allows radiological quantities establishment related to that spectrum but it is necessary, however, a series of steps with a spectrometric system that can cover a large interval of energy and whose answer is isotropic. The most widely used for accomplishing this task is the spectrometric Bonner spheres system. One of the biggest problems related to neutron spectrometry is the process of data analysis, known as unfolding. Most of the work undertaken to implement new techniques of this process, using data obtained with the scintillator {sup 6}LiI(I). However, characteristics related to the dead time make it not be so effective when used in high flow neutron fields. An alternative to this problem is the use of thermoluminescent detectors (TLD), but the codes used do not provide a more specific response matrix to unfolding the information obtained through these materials, which makes the development of a specific response matrix important to adequately characterize the response obtained by them. This paper proposes using a technique of artificial intelligence called genetic algorithm, which uses bio-inspired mathematical models and through the implementation of a specific matrix to unfolding data obtained from a combination of TLDs embedded in a system of Bonner spheres, such as thermal neutron detectors, to characterize the neutron spectrum as a function of energy. The results obtained with this method were in accordance with reference spectra, thus enables of this technique to unfolding neutrons spectra with BS-TLD system. (author)

films or films with magnetic properties. The reason is the peculiar property of neutron light since the mass of a neutron is close to the one of a proton, and since it bears a magnetic moment. The optical properties of matter, when interacting with neutrons, are described by a refractive index......Neutron (and X-ray) reflectometry constitute complementary interfacially sensitive techniques that open access to studying the structure within thin films of both soft and hard condensed matter. Film thickness starts oxide surfaces on bulk substrates, proceeding to (pauci-)molecular layers and up...... to hundreds of nanometers. Thickness resolution for flat surfaces is in the range of few Ǻngstrøm, and as a peculiar benefit, the presence and properties of buried interfaces are accessible. Focus here will be on neutron reflectometry, a technique that is unique in applications involving composite organic...

We describe the results obtained for the characterization of neutron and photon fields of the graphite thermal column of a 10-MW MTR-type research reactor, together with the mandatory Verification and Validation (V and V) procedure. The graphite thermal column exhibits a relatively small cross sectional are (50 cm X 50 cm) and a large depth ({approx}140cm), representing a difficult deep-penetration problem. The calculation line relied on estimations made the Monte Carlo MCNP-4.2 code. The Validation and Verification (V and V) procedure required: mandatory norms and an auditable path; ANISN/VITAMIN-C (deterministic) calculations for comparison with MCNP; identification of all approximations used, together with the method of justification; explicit statement of parameters for comparison, and statement of the areas of applicability; parametric studies concerning impurities and transverse leakage effects; statement of biases and uncertainties. The results showed a restricted range of applicability ({approx}60 cm) for fast and epithermal neutronfluxes, therefore needing a careful extrapolation to deeper locations. The transverse leakage of neutrons has a greater effect on the diffusion of thermal neutrons when compared to impurities of up to 5 ppm of natural boron. In addition, it is discussed the nature of the biases and uncertainty bands calculated. (author)

This report describes animal dosimetry studies and phantom measurements. During 1994, 12 dogs were irradiated at BMRR as part of a 4 fraction dose tolerance study. The animals were first infused with BSH and irradiated daily for 4 consecutive days. BNL irradiated 2 beagles as part of their dose tolerance study using BPA fructose. In addition, a dog at WSU was irradiated at BMRR after an infusion of BPA fructose. During 1994, the INEL BNCT dosimetry team measured neutronflux and gamma dose profiles in two phantoms exposed to the epithermal neutron beam at the BMRR. These measurements were performed as a preparatory step to the commencement of human clinical trials in progress at the BMRR.

The thermal-neutronflux depression within a fuel rod has a great influence in the radial temperature profile of the rod, especially for high enrichment fuel. For this reason, a study was made about the UO{sub 2}-PUO{sub 2} (15 to 30% PUO{sub 2}) fuel pins for the KfK-JEN joint irradiation program IVO, in the FR2 reactor. Different methods (diffusion, Bonalumi, successive generations) were compared and a new approach (parabolic approximation) was developed. (Author) 22 refs.

The spectral neutron and photon fluence (or flux) measured outside and inside of assemblies related to fusion reactor constructions are basic quantities of fusion neutronics. The comparison of measured spectra with the results of MCNP neutron and photon transport calculations allows a crucial test of evaluated nuclear data as generally used in fusion applications to be carried out. The experiments concern mixed neutron/photon fields with about the same intensity of the two components. An NE-213 scintillation spectrometer, well described by response matrices for both neutrons and photons, is used as proton-recoil and Compton spectrometer. The experiments described here in more detail address the background problematic of two applications, an iron benchmark experiment with an ns-pulsed neutron source and a deep penetration mock-up experiment for the investigation of the ITER in-board shield system. The measured spectral neutron and photon fluences are compared with spectra calculated with the MCNP code on the b...

For the research of CCD neutron radiography, a neutron collimator was designed based on the exit of thermal neutron of the Boron Neutron Capture Therapy (BNCT) reactor. Based on the Geant4 simulations, the preliminary choice of the size of the collimator was determined. The materials were selected according to the literature data. Then, a collimator was constructed and tested on site. The results of experiment and simulation show that the thermal neutronflux at the end of theneutron collimator is greater than 10^6 n/cm^2/s, the maximum collimation ratio (L/D) is 58, the Cd-ratio(Mn) is 160 and the diameter of collimator end is 10 cm. This neutron collimator is considered to be applicable for neutron radiography.

While the neutrons within the core may take part in the chain reaction, those neutrons emitted from the core are basically lost for the energy production. This 'neutron leakage' represents a loss of fuel efficiency and causes neutron embrittlement of the reactor pressure vessel (RPV) wall. The latter raises safety concerns, needs to be monitored closely and may necessitate mitigating measures. There are different strategies to deal with these two undesirable effects: The neutron emission may be reduced to some extent all around the core or just at the 'hot spots' of RPV embrittlement by tailored core loading patterns. A higher absorption rate of neutrons may also be achieved by a larger water gap between the core and the RPV. In this paper the inter-relations between the distribution of neutronflux, core geometry, core loading strategy, RPV embrittlement and its surveillance are discussed at first. Then the different strategies followed by the German and French operators are described. Finally the conclusions will highlight the communalities and differences between these strategies as different approaches to the same problem of safety as well as economy. (authors)

This paper summarizes the work, performed in the frame of various EFDA contracts during 2004-2005, on the design review and upgrade of the ITER radial neutron camera (RNC). The RNC, which should provide information on the spatial distribution and energy spectrum of the neutron emission, consists of an ex-vessel system (fan-like collimator with 12 x 3 lines of sights) and an in-vessel system with further 9 lines for a full coverage of the plasma. A Monte Carlo code (MCNP) has been used for the neutronic calculations. The basic ITER model has been developed from the CATIA drawings to include the RNC with all details relevant for the neutronic analysis. In the model the collimator diameters have been set to 2 and 4 cm, respectively, for the ex-vessel and in-vessel systems. A detailed space dependent fusion neutron source (DD and DT phases in various plasma scenarios) has been used with a consistent ion temperature radial profile. A special variance reduction treatment has been developed so that neutrons reach the far regions in the high collimated neutron beam and score with a satisfying statistical error. Neutron and photon fluxes and spectra have been calculated. Approximately, one neutron out of 10{sup 11} emitted in all the plasma reaches a single ex-vessel detector. Therefore, for an emission rate of 1.8 x 10{sup 20} n/s (corresponding to 500 MW fusion power) the flux on the detectors is in the range (1-5) x 10{sup 8} n/(cm{sup 2} s) depending on the poloidal orientation. The fraction of scattered neutrons (>1 MeV) is lower than few % of the total. A measurement simulation software tool (MSST) performing asymmetric Abel inversion of simulated measured neutron signals has also been developed for line of sight and design optimization. Combining information from MCNP calculations and MSST, it has been possible to evaluate the performance of the RNC, check whether the present design of the RNC meets the measurement requirements and optimize the RNC design.

Using a {sup 252}Cf neutron source we have calibrated the response of Ge detectors to fast neutrons and deduced the expression for determination of the flux of fast environmental neutrons based on the intensity of the spectral structure at 691.3 keV. We also suggest an approximate expression for estimation of the flux of environmental thermal neutrons from the intensity of the 139.9 keV line. (orig.).

We discuss the principle and performance of a new cold-neutron bender using sequential garland reflections in order to bend a neutron beam with a large divergence to a large angle. Using this bender for a pulsed neutron source, we can distribute cold neutrons or polarized cold neutrons if necessary to plural spectrometers at a cold-neutron-guide tube to avoid the frame-overlap problem of cold neutrons. (orig.)

The `whispering gallery' effect has been known since ancient times for sound waves in air, later in water and more recently for a broad range of electromagnetic waves: radio, optics, Roentgen and so on. It consists of wave localization near a curved reflecting surface and is expected for waves of various natures, for instance, for atoms and neutrons. For matter waves, it would include a new feature: a massive particle would be settled in quantum states, with parameters depending on its mass. Here, we present for the first time the quantum whispering-gallery effect for cold neutrons. This phenomenon provides an example of an exactly solvable problem analogous to the `quantum bouncer'; it is complementary to the recently discovered gravitationally bound quantum states of neutrons . These two phenomena provide a direct demonstration of the weak equivalence principle for a massive particle in a pure quantum state. Deeply bound whispering-gallery states are long-living and weakly sensitive to surface potential; highly excited states are short-living and very sensitive to the wall potential shape. Therefore, they are a promising tool for studying fundamental neutron-matter interactions, quantum neutron optics and surface physics effects.

For nuclear data measurements with a white neutron source, to control the background at the detector is a key issue. The neutron dump which locates at the end of the white neutron beam line at CSNS has a very important impact to the neutron and gamma backgrounds in the endstation. A sophisticated neutron dump was designed to reduce the backgrounds to the level of about 10-8 relative to the neutronflux. In this paper, the method to suppress both neutron and gamma backgrounds near a white-spectrum neutron dump is introduced. The optimized geometry structure and materials of the dump are described, and the neutron and gamma space distributions have been calculated by using the FLUKA code for different operation settings which are defined by beam spots of Φ30 mm, Φ60 mm and 90 mm×90 mm, respectively.

Neutron spectrometry provides many branches of science and technology with the necessary data. Usually the main part of the data is supplied by powerful neutron time-of-ﬂight spectrometers. Nevertheless there are many other very effective but simpler and cheaper neutron spectroscopy methods on accelerators, suitable for solution of plenty of scientiﬁc and applied problems (for example, in astrophysics and radioactive waste transmutation). The methods of slowing-down spectrometry in lead and graphite, generating of neutron spectra, characteristic for nucleosynthesis in the stars, and neutron spectrometry by means of primary -transition shift are discussed in the report.

ITER is an experimental nuclear reactor, aiming to demonstrate the feasibility of nuclear fusion realization in order to use it as a new source of energy. ITER is a plasma device (tokamak type) which will be equipped with a set of plasma diagnostic tools to satisfy three key requirements: machine protection, plasma control and physics studies by measuring about 100 different parameters. ITER diagnostic equipment is integrated in several ports at upper, equatorial and divertor levels as well internally in many vacuum vessel locations. The Diagnostic Systems will be procured from ITER Members (Japan, Russia, India, United States, Japan, Korea and European Union) mainly with the supporting structures in the ports. The various diagnostics will be challenged by high nuclear radiation and electromagnetic fields as well by severe environmental conditions (ultra high vacuum, high thermal loads). Several neutron systems with different sensitivities are foreseen to measure ITER expected neutron emission from 1014 up to almost 1021 n/s. The measurement of total neutron emissivity is performed by means of NeutronFlux Monitors (NFM) installed in diagnostic ports and by Divertor NeutronFlux Monitors (DNFM) plus MicroFission Chambers (MFC) located inside the vacuum vessel. The neutron emission profile is measured with radial and vertical neutron cameras. Spectroscopy is accomplished with spectrometers looking particularly at 2.5 and 14 MeV neutron energy. Neutron Activation System (NAS), with irradiation ends inside the vacuum vessel, provide neutron yield data. A calibration strategy of the neutron diagnostics has been developed foreseeing in situ and cross calibration campaigns. An overview of ITER neutron diagnostic systems and of the associated challenging engineering and integration issues will be reported.

Rhenium-188, a daughter product of tungsten-188, is an isotope of great interest in therapeutic nuclear medicine, being used in dozens of laboratory and clinical investigations worldwide. Applications include various cancer therapy strategies, treatment of rheumatoid arthritis, prevention of restenosis following coronary artery angioplasty, and palliation of bone pain associated with cancer metastases. With its half-life of 17 hours, 2.12 MeV (maximum) beta-particle emission, chemical similarity to technetium-99m (the most widely used diagnostic radioisotope), and its availability in a convenient tungsten-188/rhenium-188 generator system, rhenium-188 is a superb candidate for a broad range of applications. Production of 188W is typically via double neutron capture by 186W in a high flux nuclear reactor, predominantly the High Flux Isotope Reactor at the Oak Ridge National Laboratory in Tennessee. Experience at HFIR has shown that production yields (measured in Ci of 188W produced per g of 186W target) decrease considerably as target size increases. While the phenomenon of neutron resonance self-shielding would be expected to produce such an effect, temperature effects on neutronflux distribution and neutron capture rates may also be involved. Experimental investigations of these phenomena have not been previously performed. The work presented in this thesis evaluates the factors that contribute to the decrease in 188W yield from both theoretical and experimental standpoints. Neutron self-shielding and temperature effects were characterized to develop a strategy for target design that would optimize production yield, an important factor in minimizing health care costs. It was determined that decrease in yield due to neutron self-shielding can be attributed to depletion of epithermal neutrons at resonant energies, most significantly within the initial 0.4 mm depth of the target. The results from these studies further show that 188W yield in the interior of the

We report measurements of the flux of fast neutrons at Mercury from 20ºS to the north pole. On the basis of neutron transport simulations and remotely sensed elemental compositions, cosmic-ray-induced fast neutrons are shown to provide a measure of average atomic mass, , a result consistent with earlier studies of the Moon and Vesta. The dynamic range of fast neutronflux at Mercury is 3%, which is smaller than the fast-neutron dynamic ranges of 30% and 6% at the Moon and Vesta, respectively. Fast-neutron data delineate compositional terranes on Mercury that are complementary to those identified with X-ray, gamma-ray, and slow-neutron data. Fast neutron measurements confirm the presence of a region with high , relative to the mean for the planet, that coincides with the previously identified high-Mg region and reveal the existence of at least two additional compositional terranes: a low- region within the northern smooth plains and a high- region near the equator centered near 90ºE longitude. Comparison of the fast-neutron map with elemental composition maps show that variations predicted from the combined element maps are not consistent with the measured variations in fast-neutronflux. This lack of consistency could be due to incomplete coverage for some elements or uncertainties in the interpretations of compositional and neutron data. Currently available data and analyses do not provide sufficient constraints to resolve these differences.

Following the shutdown of FRJ-2 research reactor in Jülich, the pinhole small-angle neutron diffractometers KWS-1 and KWS-2 have been moved to the research reactor FRM II in Garching. The installation of these 40 m long instruments required the design and setup of new neutron guides with geometrical and optical features imposed by the instruments' positioning in the neutron guide hall, such as, the predetermined length and beam height as well as the foreseen improvement of the instrument performance. We report here about the quality assessment of the newly constructed neutron guides with respect to the optical, geometrical and alignment characteristics and the positioning of the velocity selector integrated in the neutron guide system by comparing the features of the measured neutron beams (in terms of neutronflux, intensity distribution and beam profile) with the results of the simulations of optimal neutron guide systems.

High-fluxneutron sources for neutron imaging and materials analysis applications have typically been provided by accelerator-based (Spallation Neutron Source) and reactor-based (High Flux Isotope Reactor) neutron sources. A novel approach is to use ultra-intense (> 10^18 W/cm^2) laser-target interactions to generate picosecond, collimated neutrons. Here we examine the feasibility of a source based on current (LULI) and upcoming laser facility capabilities. A Monte-Carlo code calculates angular and energy distributions of neutrons generated by D-D fusion events occurring within a deuterated target for a given incident beam of D+ ions. The parameters of the deuteron beam are well understood from laser-plasma and laser-target studies relevant to fast-ignition fusion. Expected neutron yields are presented in comparison to conventional neutron sources, previous experimental neutron yields, and within the context of neutron shielding safety requirements.

The familiar extremes of crystalline material are single-crystals and random powders. In between these two extremes are polycrystalline aggregates, not randomly arranged but possessing some preferred orientation and this is the form taken by constructional materials, be they steel girders or the bones of a human or animal skeleton. The details of the preferred orientation determine the ability of the material to withstand stress in any direction. In the case of bone the crucial factor is the orientation of the c-axes of the mineral content - the crystals of the hexagonal hydroxyapatite - and this can readily be determined by neutron diffraction. In particular it can be measured over the volume of a piece of bone, utilizing distances ranging from 1mm to 10mm. The major practical problem is to avoid the intense incoherent scattering from the hydrogen in the accompanying collagen; this can best be achieved by heat-treatment and it is demonstrated that this does not affect the underlying apatite. These studies of bone give leading anatomical information on the life and activities of humans and animals - including, for example, the life history of the human femur, the locomotion of sheep, the fracture of the legs of racehorses and the life-styles of Neolithic tribes. We conclude that the material is placed economically in the bone to withstand the expected stresses of life and the environment. The experimental results are presented in terms of the magnitude of the 0002 apatite reflection. It so happens that for a random powder the 0002, 1121 reflections, which are neighboring lines in the powder pattern, are approximately equal in intensity. The latter reflection, being of manifold multiplicity, is scarcely affected by preferred orientation so that the numerical value of the 0002/1121 ratio serves quite accurately as a quantitative measure of the degree of orientation of the c-axes in any chosen direction for a sample of bone.

The detector materials were selected and proved for the thermal, intermediate and quick energy intervals of the neutron spectra, in the radial tube RW-2 to 1 MW of thermal power during 5 min, being obtained good results respect to the activation of the thin sheets. However, since the exhibition speed, of each experimental arrangement after it irradiation that was of the order of 2 R/h, it was considered that may be more convenient the irradiation, separated from each thin sheet, for that which it was selected the SINCA irradiation system, because it simplified the activities related with the individual irradiation of the thin sheets and the count of the same ones. It was found that the selection, the geometric arrangement and measurement of the activities of the used thin sheets, its are the three factors that more affect the input data of the SAND II code that it will be used in our case for the determination of the neutron spectra of the Reactor. (Author)

A transient flux of high energy solar neutrons from 50 MeV to about 1 GeV has been detected by the Gamma Ray Spectrometer (GRS) on the Solar Maximum Mission (SMM) satellite following an intense burst of high energy photons (less than 100 MeV) peaking at 1143:29 UT. The neutrons were also detected by the IGY neutron monitor on Jungfraujoch (Switzerland). In this paper the SMM GRS observations are summarized and compared with the Jungfraujoch neutron monitor data, and both the time dependent neutronflux at the earth and the neutron emission spectrum at the sun are estimated.

We present recent developments in fast neutron detection using segmented spectrometers based on the principle of capture-gating. Our approach employs an organic scintillator to detect fast neutrons through their recoil interaction with protons in the scintillator. The neutrons that thermalize and are captured produce a signal indicating that the event was due to a neutron recoil and that the full energy of the neutron was deposited. The delayed neutron capture also serves to discriminate against uncorrelated background events. The segmentation permits reconstruction of the initial neutron energy despite the nonlinear response of the scintillator. We have constructed spectrometers using both He-3 proportional counters and Li-6 doping as capture agents in plastic and liquid organic scintillators. We discuss the operation of the spectrometers for the measurement of low levels of fast neutrons for several applications, including the detection of very low-activity neutron sources and the characterization of the flux and spectrum of fast neutrons at the Earth's surface and in the underground environment.

Kyoto University Research Reactor Institute (KURRI) and Sumitomo Heavy Industries, Ltd. (SHI) have developed a cyclotron-based neutron source for Boron Neutron Capture Therapy (BNCT). It was installed at KURRI in Osaka prefecture. The neutron source consists of a proton cyclotron named HM-30, a beam transport system and an irradiation & treatment system. In the cyclotron, H- ions are accelerated and extracted as 30 MeV proton beams of 1 mA. The proton beams is transported to the neutron production target made by a beryllium plate. Emitted neutrons are moderated by lead, iron, aluminum and calcium fluoride. The aperture diameter of neutron collimator is in the range from 100 mm to 250 mm. The peak neutronflux in the water phantom is 1.8×109 neutrons/cm2/sec at 20 mm from the surface at 1 mA proton beam. The neutron source have been stably operated for 3 years with 30 kW proton beam. Various pre-clinical tests including animal tests have been done by using the cyclotron-based neutron source with 10B-p-Borono-phenylalanine. Clinical trials of malignant brain tumors will be started in this year.

Kyoto University Research Reactor Institute (KURRI) and Sumitomo Heavy Industries, Ltd. (SHI) have developed a cyclotron-based neutron source for Boron Neutron Capture Therapy (BNCT). It was installed at KURRI in Osaka prefecture. The neutron source consists of a proton cyclotron named HM-30, a beam transport system and an irradiation and treatment system. In the cyclotron, H- ions are accelerated and extracted as 30 MeV proton beams of 1 mA. The proton beams is transported to the neutron production target made by a beryllium plate. Emitted neutrons are moderated by lead, iron, aluminum and calcium fluoride. The aperture diameter of neutron collimator is in the range from 100 mm to 250 mm. The peak neutronflux in the water phantom is 1.8 Multiplication-Sign 109 neutrons/cm{sup 2}/sec at 20 mm from the surface at 1 mA proton beam. The neutron source have been stably operated for 3 years with 30 kW proton beam. Various pre-clinical tests including animal tests have been done by using the cyclotron-based neutron source with {sup 10}B-p-Borono-phenylalanine. Clinical trials of malignant brain tumors will be started in this year.

The HANARO (KAERI) research reactor has been developed a neutron guide system for cold neutron (CN) research facilities since July, 2003. The neutron guide system plays an important role in transporting cold neutrons from the CN source to the neutron facilities as CN-NDP, CN-PGAA, SANS, etc. The CN activation station is being installed in the HANARO cold-neutron research project. The CN-NDP and CN-PGAA were selected as two facilities using at this station. At the end position of CG1 and CG2B beam guides, the CN-NDP and CN-PGAA will be installed in the CN guide hall. In order to predict the neutronflux and intensity values at the CG1 and CG2B beam guides, the simulation results of neutronflux at the CG1 and CG2B beam guides are presented by using several Monte Carlo (MC) neutron ray-tracing simulation codes. The intercomparison of neutronflux values between McStas, VITESS and RESTRAX are performed for getting fairly correct results at two neutron beam guides

Kamini (Kalpakkam Mini) reactor is a U233 fuelled, demineralised light water moderated and cooled, beryllium oxide reflected, low power (30 kW) nuclear research reactor. This reactor functions as a neutron source with a flux of 1012 n/cm2 s-1 at core centre with facilitates for carrying out neutron radiography, neutron activation analysis and neutron shielding experiments. There are two beam tubes for neutron radiography. The length/diameter ratio of the collimators is about 160 and the aperture size is 220 mm×70 mm. Flux at the outer end of the beam tube is ˜106-107 n/cm2 s. The north end beam tube is for radiography of inactive object while the south side beam tube is for radiography of radioactive objects. The availability of high neutronflux coupled with good collimated beam provides high quality radiographs with short exposure time. The reactor being a unique national facility for neutron radiography has been utilized in the examination of irradiated components, aero engine turbine blades, riveted plates, automobile chain links and for various types of pyro devices used in the space programme. In this paper, an overview of the salient features of this reactor facility for neutron radiography and our experience in the inspection of a variety of industrial components will be given.

The High Energy Density Physics group is interested in the basic science of creating a neutron and gamma ray source. The neutrons and gamma rays are produced by accelerating ions via a laser into a target and creating fusion neutrons and gamma rays. A scintillator and photomultiplier tube will be used to detect these neutrons. Neutrons and photons produce ionizing radiation in the scintillator which then activates metastable states. These metastable states have both short and long decay rates. The initial photon count is orders of magnitude higher than the neutron count and poses problems for accurately detecting the neutrons due to the long decay state that is activated by the photons. The effects of adding lead shielding on the temporal response and signal level of the neutron detector will be studied in an effort to minimize the photon count without significant reduction to the temporal resolution of the detector. MCNP5 will be used to find the temporal response and energy deposition into the scintillator by adding lead shielding. Results from the simulations will be shown. Optimization of our scintillator neutron detection system is needed to resolve the neutron energies and neutron count of a novel neutron and gamma ray source.

Nuclear instrumentation system is an important part of safety instrument and control system in nuclear power plant. This paper introduces the principle of AP1000 power range high positive neutronflux rate reactor trip, and analyses the relationship of the trigger of this reactor trip with setpoint, time constant and power rate. The design of setpoint and time constant for power range high positive neutronflux rate reactor trip is verified on MATLAB software platform based on the transient process data of rod ejection and normal operation transient also.%核仪表系统(RPN)是核电厂仪控系统的重要组成部分。本文介绍了AP1000功率量程中子注量率正变化率高紧急停堆的基本原理，对该紧急停堆信号触发与停堆定值、时间常数、功率变化率的关系进行了分析，并以MATLAB软件为平台，基于弹棒事故和正常运行瞬态两种典型工况的瞬态过程数据，对AP1000功率量程中子注量率正变化率高紧急停堆定值和时间常数的设计进行了仿真验证。

The limitations in neutronflux and resolution (L/D) of current neutron imaging systems can be addressed with a Coded Source Imaging system with magnification (xCSI). More precisely, the multiple sources in an xCSI system can exceed the flux of a single pinhole system for several orders of magnitude, while maintaining a higher L/D with the small sources. Moreover, designing for an xCSI system reduces noise from neutron scattering, because the object is placed away from the detector to achieve magnification. However, xCSI systems are adversely affected by correlated noise such as non-uniform illumination of the neutron source, incorrect sampling of the coded radiograph, misalignment of the coded masks, mask transparency, and the imperfection of the system Point Spread Function (PSF). We argue that a model-based reconstruction algorithm can overcome these problems and describe the implementation of a Simultaneous Iterative Reconstruction Technique algorithm for coded sources. Design pitfalls that preclude a satisfactory reconstruction are documented.

A neutronics analysis using the Monte Carlo method is carried out for the end-plug penetration and magnet system of a tandem mirror fusion reactor. Detailed penetration and the magnets' three-dimensional configurations are modeled. A method of position dependent angular source biasing is developed to adequately sample the DT fusion source in the central cell region and obtain flux contributions at the penetration components. To assure cryogenic stability, the barrier cylindrical solenoid is identified as needing substantial shielding of about 1 m of a steel-lead-boron-carbide-water mixture. Heating rates there would require a thermal-hydraulic design similar to that in the central cell blanket region. The transition coils, however, need a minimal 0.2 m thickness shield. The leakage neutronflux at the direct converters is estimated at 1.3×1015 n/(m2·s), two orders of magnitude lower than that reported at the neutral beam injectors for tokamaks around 1017 n/(m2·s) for a 1 MW/m2 14 MeV neutron wall loading. This result is obtained through a coupling between the nuclear and plasma physics designs in which hydrogen ions rather than deuterium atoms are used for energy injection at the end plug, to avoid creating a neutron source there. This lower and controllable radiation leakage problem is perceived as a potential major advantage of tandem mirrors compared to tokamaks and laser reactor systems.

The neutron spin is a critical degree of freedom for many precision measurements using low-energy neutrons. Fundamental symmetries and interactions can be studied using polarized neutrons. Parity-violation (PV) in the hadronic weak interaction and the search for exotic forces that depend on the relative spin and velocity, are two questions of fundamental physics that can be studied via the neutron spin rotations that arise from the interaction of polarized cold neutrons and unpolarized matter. The Neutron Spin Rotation (NSR) collaboration developed a neutron polarimeter, capable of determining neutron spin rotations of the order of 10-7 rad per meter of traversed material. This paper describes two key components of the NSR apparatus, responsible for the transport and manipulation of the spin of the neutrons before and after the target region, which is surrounded by magnetic shielding and where residual magnetic fields need to be below 100 μG. These magnetic field devices, called input and output coils, provide the magnetic field for adiabatic transport of the neutron spin in the regions outside the magnetic shielding while producing a sharp nonadiabatic transition of the neutron spin when entering/exiting the low-magnetic-field region. In addition, the coils are self contained, forcing the return magnetic flux into a compact region of space to minimize fringe fields outside. The design of the input and output coils is based on the magnetic scalar potential method.

Earth is connected gravitationally, magnetically and electrically to its heat source - a neutron star that is obscured from view by waste products in the photosphere. Neutron repulsion is like the hot filament in an incandescent light bulb. Excited neutrons are emitted from the solar core and decay into hydrogen that glows in the photosphere like a frosted light bulb. Neutron repulsion was recognized in nuclear rest mass data in 2000 as the overlooked source of energy, the keystone of an arch that locked together these puzzling space-age observations: 1.) Excess 136Xe accompanied primordial helium in the stellar debris that formed the solar system (Fig. 1); 2.) The Sun formed on the supernova core (Fig. 2); 3.) Waste products from the core pass through an iron-rich mantle, selectively carrying lighter elements and lighter isotopes of each element into the photosphere (Figs. 3-4); and 4.) Neutron repulsion powers the Sun and sustains life (Figs. 5-7). Together these findings offer a framework for understanding...

In boron neutron capture therapy (BNCT),the ratio of the fast neutronflux to the neutronflux in the tumor (RFNT) must be less than 3％.If a D-T neutron generator is used in BNCT,the 14 MeV neutron moderator must be optimized to reduce the RFNT.Based on the neutron moderation theory and the simulation results,tungsten,lead and diamond were used to moderate the 14 MeV neutrons.Satisfying RFNT of less than 3％,the maximum neutronflux in the tumor was achieved with a three-layer moderator comprised of a 3 cm thick tungsten layer,a 14 cm thick lead layer and a 21 cm thick diamond layer.

Neutron radiography is a non-destructive testing developing fast recently, which requires stable and proper neutron source with low γ background. Neutrons from In-hospital Neutron Irradiator (IHNI) could meet this requirement. Based on the neutron beams of IHNI, a collimator is designed and built for neutron radiography. The experiment results show that in the case of IHNI working at normal rated power, the neutronflux at the end of the collimator is 1.43×106 cm-2·s-1; The max collimation ratio (L/D) is 58; the γ dose rate is 6.3×106 mSv/s. In a word, the collimator could be used for neutron radiography.

Neutrons have an important role to play in structural biology. Neutron crystallography, small-angle neutron scattering and inelastic neutron scattering techniques all contribute unique information on biomolecular structures. In particular, solution scattering techniques give critical information on the conformations and dispositions of the components of complex assemblies under a wide variety of relevant conditions. The power of these methods is demonstrated here by studies of protein/DNA complexes, and Ca{sup 2+}-binding proteins complexed with their regulatory targets. In addition, we demonstrate the utility of a new structural approach using neutron resonance scattering. The impact of biological neutron scattering to date has been constrained principally by the available fluxes at neutron sources and the true potential of these approaches will only be realized with the development of new more powerful neutron sources. (author)

NECTAR (Neutron Computerized Tomography and Radiography) is a versatile facility for radiographic and tomographic investigations as well as for neutron activation experiments using fission neutrons. The radiation sources for this facility are two plates of highly enriched uranium situated in the moderator vessel in FRM II. Thermal neutrons originating from the main fuel element of the reactor generate in these plates fast neutrons. These can escape through a horizontal beam tube without moderation. The beam can be filtered and manipulated in order to reduce the accompanying gamma radiation and to match the specific experimental tasks. A summary of the main parameters required for experimental set-up and (quantitative) data evaluation is presented. The (measured) spectra of the neutron and gamma radiations are shown along with the effect of different filters on their behavior. The neutron and gamma fluxes, dose rates, L/D-ratios, etc. and the main parameters of the actually used detection systems for neutron imaging are given, too.

A transportable fast neutron detection system has been designed and constructed for measuring neutron energy spectra and flux ranging from tens to hundreds of MeV. The transportability of the spectrometer reduces the detector-related systematic bias between different neutron spectra and flux measurements, which allows for the comparison of measurements above or below ground. The spectrometer will measure neutronfluxes that are of prohibitively low intensity compared to the site-specific background rates targeted by other transportable fast neutron detection systems. To measure low intensity high-energy neutronfluxes, a conventional capture-gating technique is used for measuring neutron energies above 20 MeV and a novel multiplicity technique is used for measuring neutron energies above 100 MeV. The spectrometer is composed of two Gd containing plastic scintillator detectors arranged around a lead spallation target. To calibrate and characterize the position dependent response of the spectrometer, a Monte Carlo model was developed and used in conjunction with experimental data from gamma ray sources. Multiplicity event identification algorithms were developed and used with a Cf-252 neutron multiplicity source to validate the Monte Carlo model Gd concentration and secondary neutron capture efficiency. The validated Monte Carlo model was used to predict an effective area for the multiplicity and capture gating analyses. For incident neutron energies between 100 MeV and 1000 MeV with an isotropic angular distribution, the multiplicity analysis predicted an effective area of 500 cm{sup 2} rising to 5000 cm{sup 2}. For neutron energies above 20 MeV, the capture-gating analysis predicted an effective area between 1800 cm{sup 2} and 2500 cm{sup 2}. The multiplicity mode was found to be sensitive to the incident neutron angular distribution.

High-energy x-ray radiotherapy machines generate neutrons by photonuclear reactions in the target and the treatment head and expose the patient to a neutronflux. In order to evaluate the neutron exposure quantitatively, fast and thermal neutron profiles for 25-MV x-ray beams of the Sagittaire accelerator have been measured. An activation technique, using the reactions 31P(n, gamma)32P (thermal neutrons) and 31P(n, p)31Si (fast neutrons, E greater than 0.7 MeV), has been developed to measure fast- and thermal-neutronfluxes in an intense high-energy photon flux. The sensitivity of this activation detector to high-energy photons, which has plagued many previous neutron measurements, was carefully measured and found to be less than 4%. Neutronfluxes for various photon field sizes ranging from 5 X 5 cm to 30 X 30 cm have been measured. The fast-neutron profiles were observed to have rounded edges and the thermal fluxes were found to be relatively uniform. In the central part of the x-ray beam, the ratio of neutron dose equivalent to photon absorbed dose was found to be between 0.2% and 0.5%. Outside of the photon field, the ratio of neutron dose equivalent to the central-axis photon absorbed dose was 0.12%.

Neutron imaging has a broad scope of applications and has played a pivotal role in visualizing and quantifying hydrogenous masses in metallic matrices. The field continues to expand into new applications with the installation of new neutron imaging facilities. In this scope, a neutron imaging facility for computed tomography and real-time neutron radiography is currently being developed around 2.0MW TRIGA MARK-II reactor at Maamora Nuclear Research Center in Morocco (Reuscher et al., 1990 [1]; de Menezes et al., 2003 [2]; Deinert et al., 2005 [3]). The neutron imaging facility consists of neutron collimator, real-time neutron imaging system and imaging process systems. In order to reduce the gamma-ray content in the neutron beam, the tangential channel was selected. For power of 250 kW, the corresponding thermal neutronflux measured at the inlet of the tangential channel is around 3×10 11 ncm 2/s. This facility will be based on a conical neutron collimator with two circular diaphragms with diameters of 4 and 2 cm corresponding to L/D-ratio of 165 and 325, respectively. These diaphragms' sizes allow reaching a compromise between good flux and efficient L/D-ratio. Convergent-divergent collimator geometry has been adopted. The beam line consists of a gamma filter, fast neutrons filter, neutron moderator, neutron and gamma shutters, biological shielding around the collimator and several stages of neutron collimator. Monte Carlo calculations by a fully 3D numerical code GEANT4 were used to design the neutron beam line ( http://www.info.cern.ch/asd/geant4/geant4.html[4]). To enhance the neutron thermal beam in terms of quality, several materials, mainly bismuth (Bi) and sapphire (Al 2O 3) were examined as gamma and neutron filters respectively. The GEANT4 simulations showed that the gamma and epithermal and fast neutron could be filtered using the bismuth (Bi) and sapphire (Al 2O 3) filters, respectively. To get a good cadmium ratio, GEANT 4 simulations were used to

This paper reviews the current status of the theoretical models of the evolution of the magnetic fields of neutron stars other than magnetars. It appears that the magnetic fields of neutron stars decay significantly only if they are in binary systems. Three major physical models for this, namely spindown-induced flux expulsion, ohmic evolution of crustal field and diamagnetic screening of the field by accreted plasma, are reviewed.

Measurements of neutron emission from a fusion plasma can provide a wealth of information on the underlying temporal, spatial and energy distributions of reacting ions and how they are affected by a wide range of magneto-hydro-dynamic (MHD) instabilities. This thesis focuses on the interpretation of the experimental measurements recorded by neutronflux monitors with and without spectroscopic capabilities installed on the Mega Ampere Spherical Tokamak (MAST). In particular, the temporally and...

We present a conceptual design of a low-energy neutron generator for treatment of brain tumors by boron neutron capture theory (BNCT). The concept is based on a 2.5-MeV proton beam from a radio-frequency quadrupole (RFQ) linac, and the neutrons are produced by the /sup 7/Li(p,n)/sup 7/Be reaction. A liquid lithium target and modulator assembly are designed to provide a high flux of epithermal neutrons. The patient is administered a tumor-specific /sup 10/Be-enriched compound and is irradiated by the neutrons to create a highly localized dose from the reaction /sup 10/B(n,..cap alpha..)/sup 7/Li. An RFQ accelerator-based neutron source for BNCT is compact, which makes it practical to site the facility within a hospital. 11 refs., 5 figs., 1 tab.

The laboratory of environmental radiation of ITA (São José dos Campos, 23°11'11″S, 45°52'43″W, 650 MAMSL) performs simultaneous monitoring of a natural radiation background and meteorological parameters. Neutronflux in the energy range of 0.02 eV - 10 MeV is registered with two sets of proportional ^{3}He tubes placed into cylindrical paraffin thermalizers: an {bf outdoor detector }of 250 cm² area and {bf indoor detector }of 70 cm² area located on the second floor of a concrete building. The counter efficiency for thermal neutrons is 80%. The characteristics of the observed flux variation are quite different from those inherent to the neutrons of the cosmic ray origin. {bf Four types of the outdoor flux variations }are observed: 1) {bf seasonal }with a maxima in wet seasons; 2) {bf diurnal }with maximum at about 6 h local time and an amplitude up to several dozens; 3) {bf abrupt transient} ( 1 min) increases with magnitudes up to two orders higher than the mean daily flux; 4) short (several days) {bf quasi-periodic enhancements }with amplitudes up to several times higher than the mean daily flux. A large variation of the outdoor flux and its phase synchronism with that of the radon decay products means with a high probability their common origin. An apparent source of the neutrons observed is nuclear reactions of decay α-particles with the ground matter. In this case the dynamics of the outdoor flux variations of the first two types is controlled by those of the meteorological parameters in the locality. The third type events correlate with lightning strokes in the vicinity (<200 m) of the detector. The more rare fourth type correlate neither with geomagnetic disturbances nor with meteorological phenomena and are probably a result of natural radon release from the Earth's crust triggered by minor seismological activity. The indoor flux is quite stable with a possible weak maximum at16 h not exceeding 0.1.

Neutronics analysis, the spacer grids which support fuel rods are not explicitly described, but they are homogenized with coolant. However, the effects of neglecting or simplifying the spacer grids are not reported in the literature to the best of our knowledge. In this paper, to investigate the effects of spacer grids in neutronics analysis, a detailed description of spacer grids is added to the KAIST benchmark problem 1B. Then, the effective multiplication factor, spatial distributions of neutronflux, and its energy spectrum are obtained for the two cases (with and without spacer grids). Numerical results show that the effects of spacer grids are not negligible. In this paper, to investigate the effect of spacer grids, the spacer grid geometry is described in detail in the Monte Carlo calculation. In the numerical test, the two cases are compared in the context of a modified KAIST benchmark problem 1B. Case 1 does not have spacer grids, while the space is filled by coolant instead. Case 2 includes the spacer grids. The difference in neutronflux spectra is also observed. Thus, the effect of the spacer grids should be considered in the whole-core reactor analysis. In practice, the spacer grids are homogenized into coolant to consider its effect. As a further study, therefore, it would be worthwhile to investigate the differences between the homogenization and the explicit description of the spacer grids.

Various topics and issues on the neutronics of pulsed spallation neutron sources, mainly for neutron scattering experiments, are reviewed to give a wide circle of readers a better understanding of these sources in order to achieve a high neutronic performance. Starting from what neutrons are needed, what the spallation reaction is and how to produce slow-neutrons more efficiently, the outline of the target and moderator neutronics are explained. Various efforts with some new concepts or ideas have already been devoted to obtaining the highest possible slow-neutron intensity with desired pulse characteristics. This paper also reviews the recent progress of such efforts, mainly focused on moderator neutronics, since moderators are the final devices of a neutron source, which determine the source performance. Various governing parameters for neutron-pulse characteristics such as material issues, geometrical parameters (shape and dimensions), the target-moderator coupling scheme, the ortho-para-hydrogen ratio, po...

U.S. Geological Survey, Department of the Interior — Methane (CH4) flux is the net rate of methane exchange between an ecosystem and the atmosphere. Data of this variable were generated by the USGS LandCarbon project...